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Sample records for carolinas virginia tube reactor

  1. Pressure tube type reactor

    International Nuclear Information System (INIS)

    Heretofore, a pressure tube type reactor has a problem in that the evaluation for the reactor core performance is complicate and no sufficient consideration is made for the economical property, to increase the size of a calandria tank and make the cost expensive. Then, in the present invention, the inner diameter of a pressure tube is set to greater than 50% of the lattice gap in a square lattice like arrangement, and the difference between the inner and the outer diameters of the calandria tube is set smaller than 20% of the lattice gap. Further, the inner diameter of the pressure tube is set to greater than 40% and the difference between the inner and the outer diameters of the calandria tube is set smaller than 30% of the lattice gap in a triangle lattice arrangement. Then, heavy water-to-fuel volume ratio can be determined appropriately and the value for the coolant void coefficient is made more negative side, to improve the self controllability inherent to the reactor. In particular, when 72 to 90 fuel rods are arranged per one pressure tube, the power density per one fuel rod is can be increased by about twice. Accordingly, the number of the pressure tubes can be reduced about to one-half, thereby enabling to remarkably decrease the diameter of the reactor core and to reduce the size of the calandria, which is economical. (N.H.)

  2. Virginia Tech Corps of Cadets alumnus Shane Morton named North Carolina game Hokie Hero

    OpenAIRE

    Cox, Carrie

    2010-01-01

    Virginia Tech Corps of Cadets alumnus Cmdr. Shane Morton, U.S. Navy, who earned a degree in history from the College of Liberal Arts and Human Sciences in 1995 has been selected as the Hokie Hero for the Virginia Tech versus University of North Carolina game.

  3. Decommissioning Experience: University of Virginia Reactors, United States of America

    International Nuclear Information System (INIS)

    Full text: There were two research reactors at the University of Virginia. CAVALIER operated at a maximum power of 100 W from 1974 to 1988. The other reactor (generally referred to as the UVA reactor) operated at a maximum power of 1 MW or 2 MW from 1960 to 1998. The water contained in the pool was utilized to provide shielding during the segmenting and removal of highly contaminated components in the pool. This work was performed by divers using plasma arc cutting equipment. A cask liner was first placed in the reactor pool. The higher activity items were preferentially loaded nearest the centre of the cask, and the lower activity items loaded in the liner annulus to provide shielding. Because air sampling performed during segmentation proved that no airborne contamination was produced, no confinement structure was necessary. After shipment of the removed components, the remaining pool water was sampled and confirmed suitable for discharge to the sanitary sewer (through filters). Decontamination of the pool structures was performed using a water jet cutting process. Once the pool surfaces had been cleaned to bare concrete, surfaces were sampled for activation. The only activated concrete was detected radially around the beam tubes through the pool wall (author)

  4. Population Abundance and Genetic Structure of Black Bears in Coastal North Carolina and Virginia Using Noninvasive Genetic Techniques.

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — Master thesis on the population abundance and genetic structure of black bears in coastal North Carolina and Virginia using noninvasive genetic technigues on...

  5. Epidemiologic determinants of aural abscessation in free-living eastern box turtles (Terrapene carolina) in Virginia.

    Science.gov (United States)

    Brown, Justin D; Sleeman, Jonathan M; Elvinger, François

    2003-10-01

    Epidemiologic determinants of 46 cases of aural abscessation in free-living eastern box turtles (Terrapene carolina) admitted to the Wildlife Center of Virginia (Virginia, USA) from 1991 to 2000 were evaluated. County human population density, year and season of admission, weight, and sex did not affect the risk for box turtles to develop aural abscessation. Counties with cases of aural abscessation were not randomly distributed, but rather were clustered into two multi-county regions. Geographic location was the only risk factor associated with aural abscessation in box turtles found in this study. Possible etiologies could include chronic infectious disease, malnutrition, or chronic exposure to environmental contamination with organochlorine compounds. PMID:14733291

  6. CFD Simulation on Ethylene Furnace Reactor Tubes

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    Different mathematical models for ethylene furnace reactor tubes were reviewed. On the basis of these models a new mathematical simulation approach for reactor tubes based on computational fluid dynamics (CFD) technique was presented. This approach took the flow, heat transfer, mass transfer and thermal cracking reactions in the reactor tubes into consideration. The coupled reactor model was solved with the SIMPLE algorithm. Some detailed information about the flow field, temperature field and concentration distribution in the reactor tubes was obtained, revealing the basic characteristics of the hydrodynamic phenomena and reaction behavior in the reactor tubes. The CFD approach provides the necessary information for conclusive decisions regarding the production optimization, the design and improvement of reactor tubes, and the new techniques implementation.

  7. Documentation of Data Collection in Currituck Sound, North Carolina and Virginia, 2006-2007

    Science.gov (United States)

    Fine, Jason M.

    2008-01-01

    During 2006 and 2007, scientists from Elizabeth City State University, North Carolina Estuarine Research Reserve, the U.S. Fish and Wildlife Service, and the U.S. Geological Survey collected hydrologic and water-quality data at nine sites in and around Currituck Sound. Hydrologic and water-quality data were collected at five tributary sites--the Northwest River near Moyock, Tull Creek near Currituck, and Intracoastal Waterway near Coinjock in North Carolina, and the Albemarle and Chesapeake Canal near Princess Anne, and the North Landing River near Creeds in Virginia. In addition, data were collected at one site at the mouth of Currituck Sound (Currituck Sound at Point Harbor, North Carolina). Only water-quality data were collected at three sites in Currituck Sound and Back Bay-Currituck Sound near Jarvisburg, and Upper Currituck Sound near Corolla in North Carolina, and Back Bay near Back Bay in Virginia. The hydrologic data included water elevation and velocity, and discharge. The water-quality data included discrete samples and continuous measurements of water temperature, specific conductance, dissolved oxygen, pH, turbidity, and chlorophyll a. The hydrologic and water-quality data collected for this study were quality assured by the U.S. Geological Survey and stored in the National Water Information System database. The data collected for this project are being used to develop an unsteady multidimensional hydrodynamic and water-quality model of Currituck Sound by the U.S. Army Corps of Engineers. The purpose of this model is to provide the basis for planning and the development of best-management practices and restoration projects for Currituck Sound and its tributaries.

  8. Spray tube for reactor container

    International Nuclear Information System (INIS)

    A container of an advanced BWR reactor has a double walled steel plate comprising inner and outer shells. A spray header as a tubular pipeline constituting a spray tube is incorporated in the inside of a ceiling of the double walled steel plate (gap between the inner shell and the outer shell). A plurality of spray nozzles extend from the spray header to the inner shell of the double walled steel plate and protruded to the inside of the container. The spray header is connected to a cooling water flowing pipe secured passing through the outer shell of the double walled steel plate. The cooling water flowing pipe is connected to a pump chamber disposed to the upper portion of the container. The spray nozzles extended from the spray header are secured to the inner shell of the double walled steel plate by way of a reinforcing plate. With such a constitution, the spray header can be protected from jetting force caused upon rupture of pipelines. (I.N.)

  9. A common DLX3 gene mutation is responsible for tricho-dento-osseous syndrome in Virginia and North Carolina families.

    Science.gov (United States)

    Price, J A; Wright, J T; Kula, K; Bowden, D W; Hart, T C

    1998-10-01

    Tricho-dento-osseous syndrome (TDO) is characterised by a variable clinical phenotype primarily affecting the hair, teeth, and bone. Different clinical features are observed between and within TDO families. It is not known whether the variable clinical features are the result of genetic heterogeneity or clinical variability. A gene for TDO was localised recently to chromosome 17q21 in four North Carolina families, and a 4 bp deletion in the human distal-less 3 gene (DLX3) was identified in all affected members. A previous genetic linkage study in a large Virginia kindred with TDO indicated possible linkage to the ABO, Gc, and Kell blood group loci. To examine whether TDO exhibits genetic heterogeneity, we have performed molecular genetic analysis to determine whether affected members of this Virginia kindred have the DLX3 gene deletion identified in North Carolina families. Results show that affected subjects (n=3) from the Virginia family have the same four nucleotide deletion previously identified in the North Carolina families. A common haplotype for three genetic markers surrounding the DLX3 gene was identified in all affected subjects in the North Carolina and Virginia families. These findings suggest that all people with TDO who have been evaluated have inherited the same DLX3 gene deletion mutation from a common ancestor. The variable clinical phenotype observed in these North Carolina and Virginia families, which share a common gene mutation, suggests that clinical variability is not the result of genetic heterogeneity at the major locus, but may reflect genetic heterogeneity at other epigenetic loci or contributing environmental factors or both. PMID:9783705

  10. Influence of Cultural and Pest Management Practices on Performance of Runner, Spanish, and Virginia Market Types in North Carolina

    Directory of Open Access Journals (Sweden)

    Bridget R. Lassiter

    2016-01-01

    Full Text Available Virginia market type peanut (Arachis hypogaea L. cultivars are grown primarily in North Carolina, South Carolina, and Virginia in the US, although growers in these states often plant other market types if marketing opportunities are available. Information on yield potential and management strategies comparing these market types is limited in North Carolina. In separate experiments, research was conducted to determine response of runner, Spanish, and Virginia market types to calcium sulfate and inoculation with Bradyrhizobium at planting, planting and digging dates, planting patterns, and seeding rates. In other experiments, control of thrips (Frankliniella spp. using aldicarb, southern corn rootworm (Diabrotica undecimpunctata Howardi using chlorpyrifos, eclipta (Eclipta prostrata L. using threshold-based postemergence herbicides, and leaf spot disease (caused by the fungi Cercospora arachidicola and Cercosporidium personatum fungicide programs was compared in these market types. Results showed that management practice and market types interacted for peanut pod yield in only the planting date experiment. Yield of runner and Virginia market types was similar and exceeded yield of the Spanish market type in most experiments.

  11. Isopach grid of the modern marine sand above the top of Pleistocene surface along the inner shelf from Virginia border to Cape Hatteras, North Carolina (modsand, ESRI binary grid, 100 m cellsize, UTM Zone 18N, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  12. Isopach grid of the Quaternary sediment thickness, inner shelf and back-barrier from Virginia border to Cape Lookout, North Carolina (q0thick, ESRI binary grid, 200 m cell size, UTM Zone 18N, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  13. Nesting and pollen preference of Osmia lignaria lignaria (Hymenoptera: Megachilidae) in Virginia and North Carolina orchards.

    Science.gov (United States)

    Kraemer, M E; Favi, F D; Niedziela, C E

    2014-08-01

    Cavity-nesting megachilid bees in the genus Osmia, found throughout the Palearctic and Nearctic regions, are good candidates for domestication. In North America, Osmia lignaria Say has been reported to be an excellent pollinator of tree fruit and is currently being developed for commercial use in orchards. This is largely because of research over several decades with the western subspecies of this bee, Osmia lignaria propinqua Cresson, in western orchards. The behavior of the eastern subspecies, O. lignaria lignaria Say, in eastern orchards has not previously been reported. This study evaluated the nesting activity and pollen preference of a population of the eastern subspecies in five orchards in the foothills and piedmont regions of North Carolina and Virginia over a 2-yr period. Apple was present in all orchards and all were bordered by hardwood forest. Shelters were placed both within orchards and the forest border. Emergence dates, nest construction, and orchard bloom were monitored weekly. Bee populations increased by 2-3 times annually at most orchards. Pollen species comprising nest provisions from 720 individual nest cells were identified and quantified using scanning electron microscopy. The greatest amount of pollen (46-82%) was that of a small understory tree, Eastern redbud (Cercis canadensis L.), at all orchard sites where these trees were present nearby. The quantity of orchard pollen was relatively low, <20% at full apple bloom, except for one orchard (53%) without nearby redbud. O. lignaria lignaria appears to prefer Eastern redbud pollen over orchard pollen. PMID:24865141

  14. Human exposure to mosquito-control pesticides--Mississippi, North Carolina, and Virginia, 2002 and 2003.

    Science.gov (United States)

    2005-06-01

    Public health officials weigh the risk for mosquito-borne diseases against the risk for human exposure to pesticides sprayed to control mosquitoes. Response to outbreaks of mosquito-borne diseases has focused on vector control through habitat reduction and application of pesticides that kill mosquito larvae. However, in certain situations, public health officials control adult mosquito populations by spraying ultra-low volume (ULV) (naled, permethrin, and d-phenothrin. These ULV applications generate aerosols of fine droplets of pesticides that stay aloft and kill mosquitoes on contact while minimizing the risk for exposure to persons, wildlife, and the environment. This report summarizes the results of studies in Mississippi, North Carolina, and Virginia that assessed human exposure to ULV naled, permethrin, and d-phenothrin used in emergency, large-scale MC activities. The findings indicated ULV application in MC activities did not result in substantial pesticide exposure to humans; however, public health interventions should focus on the reduction of home and workplace exposure to pesticides. PMID:15931155

  15. Status of the University of Virginia reactor LEU conversion

    International Nuclear Information System (INIS)

    The University of Virginia began working on converting the CAVALIER and UVAR reactors to LEU fuel in the Spring of 1986. Early in 1987, based on reactor use considerations, a decision was made to shut down the CAVALIER. A decommissioning plan was submitted to the NRC, and the decommissioning order was issued in early 1992. There is now a tentative agreement to donate the CAVALIER equipment without fuel to the University of North Texas. Design calculations for the UVAR were completed, and the Safety Analysis Report was submitted to the NRC in late 1989. The DOE/EG ampersand G order to manufacture UVAR fuel was placed at B ampersand W in March 1992, and conversion is expected to take place early in 1993

  16. YouTube researcher presents future of technology in education at Virginia Tech

    OpenAIRE

    Caldwell, T. Lynn

    2008-01-01

    More than 200 graduate students, faculty, and staff gathered in Virginia Tech's Graduate Life Center on Thursday, Sept. 11, to hear Michael Wesch present on the future of technology in education and his use of YouTube in the classroom.

  17. Farm succession planning workshops scheduled for Virginia and North Carolina farmers

    OpenAIRE

    Stott, Charlie

    2005-01-01

    The average age of a Virginia farmer is now 56 years old, and a large transfer of farm assets could occur over the next two decades. To address this important issue, the Virginia Cooperative Extension has scheduled two workshops on farm business succession planning during May in eastern Virginia.

  18. A novel mycoplasma detected in association with upper respiratory disease syndrome in free-ranging eastern box turtles (Terrapene carolina carolina) in Virginia.

    Science.gov (United States)

    Feldman, Sanford H; Wimsatt, Jeffrey; Marchang, Rachel E; Johnson, April J; Brown, William; Mitchell, Joseph C; Sleeman, Jonathan M

    2006-04-01

    Clinical signs of upper respiratory tract disease-like syndrome (URTD-LS) were observed in free-ranging eastern box turtles (Terrapene carolina carolina) from Virginia, USA (May 2001-August 2003), some of which also had aural abscesses. After a Mycoplasma sp. was detected by polymerase chain reaction (PCR), a study was undertaken to better define the range of clinical signs of disease and to distinguish mycoplasma-associated URTD-LS from other suspected causes of URTD-LS and aural abscessation in box turtles. Nasal and/or ocular swabs (from turtles possessing URTD-LS) or nasal washes (from asymptomatic turtles) were collected from turtles May 2001-August 2003; samples were assayed for Mycoplasma spp., chelonian herpesvirus, and iridoviruses by PCR testing. A partial DNA sequence (933 bases) of the small ribosomal subunit (16S rRNA) of the box turtle Mycoplasma sp. was analyzed to determine its phylogenetic relatedness to other Mycoplasma spp. of veterinary interest. Mycoplasma sp. was detected in seven (six with clinical signs of URTD-LS; one asymptomatic) of 23 fortuitously collected animals from six of 11 Virginia counties. Clinical signs in Mycoplasma sp.-infected animals included unilateral to bilateral serous to mucopurulent nasal discharge, epiphora, ocular edema, and conjunctival injection. Five Mycoplasma sp.-positive animals possessed aural abscesses; two did not. Analysis of the mycoplasma 16S rRNA gene sequence from one asymptomatic and three symptomatic animals representing four counties revealed a consensus Mycoplasma sp. sequence closely related to, but distinct from, M. agassizii. None of the samples collected contained viral DNA of chelonian herpesviruses or invertebrate and vertebrate (including FV3) iridoviruses. In conclusion, a new Mycoplasma sp. was associated with URTD-LS in native box turtles from Virginia that was not codetected with other suspected causes of chelonian upper respiratory disease; there was no proof of a direct relationship

  19. Anatomy of a shoreface sand ridge revisted using foraminifera: False Cape Shoals, Virginia/North Carolina inner shelf

    Science.gov (United States)

    Robinson, Marci M.; McBride, Randolph A.

    2008-01-01

    Certain details regarding the origin and evolution of shelf sand ridges remain elusive. Knowledge of their internal stratigraphy and microfossil distribution is necessary to define the origin and to determine the processes that modify sand ridges. Fourteen vibracores from False Cape Shoal A, a well-developed shoreface-attached sand ridge on the Virginia/North Carolina inner continental shelf, were examined to document the internal stratigraphy and benthic foraminiferal assemblages, as well as to reconstruct the depositional environments recorded in down-core sediments. Seven sedimentary and foraminiferal facies correspond to the following stratigraphic units: fossiliferous silt, barren sand, clay to sandy clay, laminated and bioturbated sand, poorly sorted massive sand, fine clean sand, and poorly sorted clay to gravel. The units represent a Pleistocene estuary and shoreface, a Holocene estuary, ebb tidal delta, modern shelf, modern shoreface, and swale fill, respectively. The succession of depositional environments reflects a Pleistocene sea-level highstand and subsequent regression followed by the Holocene transgression in which barrier island/spit systems formed along the Virginia/North Carolina inner shelf not, vert, ~5.2 ka and migrated landward and an ebb tidal delta that was deposited, reworked, and covered by shelf sand.

  20. Anatomy of a shoreface sand ridge revisited using foraminifera: False Cape Shoals, Virginia/North Carolina inner shelf

    Science.gov (United States)

    Robinson, M.M.; McBride, R.A.

    2008-01-01

    Certain details regarding the origin and evolution of shelf sand ridges remain elusive. Knowledge of their internal stratigraphy and microfossil distribution is necessary to define the origin and to determine the processes that modify sand ridges. Fourteen vibracores from False Cape Shoal A, a well-developed shoreface-attached sand ridge on the Virginia/North Carolina inner continental shelf, were examined to document the internal stratigraphy and benthic foraminiferal assemblages, as well as to reconstruct the depositional environments recorded in down-core sediments. Seven sedimentary and foraminiferal facies correspond to the following stratigraphic units: fossiliferous silt, barren sand, clay to sandy clay, laminated and bioturbated sand, poorly sorted massive sand, fine clean sand, and poorly sorted clay to gravel. The units represent a Pleistocene estuary and shoreface, a Holocene estuary, ebb tidal delta, modern shelf, modern shoreface, and swale fill, respectively. The succession of depositional environments reflects a Pleistocene sea-level highstand and subsequent regression followed by the Holocene transgression in which barrier island/spit systems formed along the Virginia/North Carolina inner shelf ???5.2 ka and migrated landward and an ebb tidal delta that was deposited, reworked, and covered by shelf sand.

  1. The Relationship between Habitat Structure and Small Mammal Communities in Southeastern Virginia and Northeastern North Carolina

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — The objectives of this study were: 1 to evaluate the small mammal communities in the range of natural habitats present in southeastern Virginia and northeastern...

  2. Advanced heavy water reactor pressure tube-easy replaceability

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) is a 300 MWe vertical pressure tube type reactor. A coolant channel consists of pressure tube, made of Zr-2.5 % Nb, which is separated from cold calandria tube using garter spring spacers. The principal function of pressure tube is to support and locate the fuel assembly and allows light water coolant through fuel assembly by natural circulation. Since AHWR is designed for life of 100 years, it necessitates the replacement of pressure tubes during service life. Easy replaceability of pressure tube, along with surveillance requirements, has major bearing on the design of coolant channel assembly. The several systems and tools have been conceptualised to cater the needs for easy and quick replacement of a pressure tube during reactor shut down. This paper gives the highlights of the innovative design features of coolant channel, preliminary design and pre-requisites for replacement, and experimental programme for demonstration of easy replaceability. (author)

  3. PROGRESS IN PROCESS INTENSIFICATION: SYNTHESIS OF IMIDAZOLE DERIVATIVES USING A SPINNING TUBE-IN-TUBE REACTOR

    Science.gov (United States)

    The high purity, high throughput synthesis of a number of imidazole derivatives using a spinning tube-in-tube reactor (STT®, Kreido Laboratories, Camarillo California) has been carried out. The STT® reactor allows the high throughput production of high purity imidazole derivativ...

  4. EPOXIDATION OF SMALL ORGANIC MOLECULES USING A SPINNING TUBE-IN-TUBE REACTOR

    Science.gov (United States)

    The commodity-scale epoxidation of several organic molecules has been carried out using a Spinning Tube-in-Tube (STTr) reactor (manufactured by Kreido Laboratories). This reactor, which embodies and facilitates the use of Green Chemistry principles and Process Intensification, a...

  5. PROGRESS IN PROCESS INTENSIFICATION: SYNTHESIS OF IMINES USING A SPINNING TUBE-IN-TUBE REACTOR

    Science.gov (United States)

    The high purity, high throughput synthesis of a number of imines (Schiff bases) using a spinning tube-in-tube reactor (STT, Kreido Laboratories, Camarillo, CA) has been carried out. The STT reactor allows the high throughput production of high purity imines from a wide variety of...

  6. The pressure tubes in the CANDU power reactor

    International Nuclear Information System (INIS)

    Nuclear power reactors using zirconium alloy pressure tubes generate electricity in several countries. In Ontario CANDU reactors generate about 30 percent of the electricity produced in the province. The pressure tubes of the first five CANDU reactors were made of cold-worked Zircaloy-2, an alloy of zirconium and tin developed by the US Navy. In 1958 the USSR published information on a Zr-2.5 wt percent Nb alloy, in which the Nb promotes stabilization of the β phase, thus presenting opportunities of exploiting metallurgically strong pressure tubes analogous to the heat-treatable α-β titanium alloys. After two reactors using Zr-2.5 wt percent Nb in a quenched and aged condition were constructed, an extensive development program on cold-worked Zr-2.5 wt percent Nb pressure tubes resulted in their becoming the reference tubes for all future CANDU reactors. Pressure tubes of Zr-3.3 wt percent Sn-0.8 wt percent Nb-0.8 wt percent Mo (Excel) are in an advanced state of development. These tubes will be used in an annealed condition; projections show that they will have improved dimensional stability over the lifetime of the reactors. These improvements result from experimental programs leading to an understanding of the relationship between microstructures and fabrication variables and effects of the environment during service in nuclear reactors. (author)

  7. Guide-tube and control rod for nuclear reactor

    International Nuclear Information System (INIS)

    The inside of the nuclear reactor guide tubes is of square cross section. A control rod drives a sliding cage and control fingers. Said sliding cage carries a system of interconnected mobile spacers for guiding control fingers

  8. Ultrasonic measurement method of calandria tube sagging in CANDU reactor

    International Nuclear Information System (INIS)

    Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor (calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the calandria tube (made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, it possibly contacts with liquid injection nozzle crossing beneath the calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, calandria tube and liquid inject ion tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here

  9. Innovative Pressure Tube Light Water Reactor with Variable Moderator Control

    International Nuclear Information System (INIS)

    The features of a reactor based on multiple pressure tubes, rather than a single pressure vessel, provide the reactor with considerable flexibility for continuous design improvements and developments. This paper presents the development of innovative pressure tube light water reactor, which has the ability to advance the current pressure tubes reactors. The proposed design is aimed to simplify the pressure tubes reactors by: - replacing heavy water by a light water as a coolant and moderator, - adopting batch refueling instead of on-line refueling. Furthermore, the design is based on proven technologies, existing fuel and structure materials. Therefore, it is reasonable to expect significant capital cost savings, short licensing and introduction period of the proposed concept into the power production grid. The basic novelty of the proposed design is based on an idea of variable moderator content in the core and 'breed and burn' mode of operation. Both concepts were extensively investigated and reported in the past (2) (3) (4). In order to evaluate a practical reactor design build on proven technology, several features of the advanced CANDU reactor (ACR-1000) were adopted. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The proposed design is basically pressure tube light water reactor with variable moderator Control (PTVM LWR). This paper presents a detailed description of the PTVM core design and demonstrates the reactivity control and the 'breed and burn' mode of operation, which are implemented by the variation of the moderator in the core, from a

  10. Status of special reactor process tube loadings, November 1, 1965

    Energy Technology Data Exchange (ETDEWEB)

    Bown, R.W.

    1965-11-10

    This report gives the status of production test control tube loadings in reactor process tubes containing significant amounts of SS materials. Data are given in table form. For further description of column headings and the current discharge goal exposure plan refer to Document RL-REA-837.

  11. Structure grid of the depth to the Pleistocene surface (Q30), inner shelf and back-barrier from Virginia border to Cape Lookout, North Carolina (q30depth, ESRI binary grid, 200 m cell size, UTM Zone 18N, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  12. Grid of the thickness of sediment above the Pleistocene surface Q50, inner shelf and back-barrier from Virginia border to Cape Lookout, North Carolina (q50thick, ESRI binary grid, 200 m cell size, UTM Zone 18N, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  13. Structure grid of the depth to the Pliocene surface (Q0), inner shelf and back-barrier from Virginia border to Cape Lookout, North Carolina (q0depth,ESRI binary grid, 200 m cell size, UTM Zone 18N, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  14. Structure grid of the depth to the Pleistocene surface (Q50), inner shelf and back-barrier from Virginia border to Cape Lookout, North Carolina (q50depth, ESRI binary grid, 200 m cell size, UTM Zone 18N, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  15. Grid of the thickness of sediment above the Pleistocene surface Q30, inner shelf and back-barrier from Virginia border to Cape Lookout, North Carolina (q30thick, ESRI binary grid, 200 m cell size, UTM, Zone 18N, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  16. Structure grid of the depth to the top of Pleistocene (Q99), inner shelf and back-barrier from Virginia border to Cape Lookout, North Carolina (q99depth, ESRI binary grid, 400 m cell size, UTM Zone 18N, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  17. Reactor physics studies for a pressure tube supercritical water reactor (PT-SCWR)

    International Nuclear Information System (INIS)

    Preliminary lattice physics and full core neutronic analysis have been performed for the pressure-tube supercritical water reactor (PT-SCWR). Current CANDU reactor physics codes (WIMS-AECL and RFSP) were used for modeling this reactor. A key challenge in the physics design of this reactor is the optimization of lattice parameters to achieve the appropriate balance between coolant void reactivity (CVR) and fuel utilization. A vertically-oriented, batch-fuelled reactor is considered, with an insulated pressure tube to accommodate the high coolant temperatures and pressures. The analysis shows the reactor physics conceptual feasibility of the design, although further optimization is required. (author)

  18. Fuel rod bundles proposed for advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACRTM-1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

  19. Improving the calandria tubes for CANDU reactors

    International Nuclear Information System (INIS)

    The Zircaloy-2 calandria tube has been improved to guard against abnormal operating conditions. It has been strengthened by either thickening or eliminating the weld to withstand the consequences of a pressure tube rupture. To exploit the moderator as a heat sink, both surfaces have been roughened and the inside surface ridged to maximise heat-transfer from an over-heated fuel channel during a postulated loss of coolant accident. (author)

  20. Detailed Sections from Auger Holes in the Emporia 1:100,000-Scale Quadrangle, North Carolina and Virginia

    Science.gov (United States)

    Weems, Robert E.; Schindler, J. Stephen; Lewis, William C.

    2010-01-01

    The Emporia 1:100,000-scale quadrangle straddles the Tidewater Fall Line in southern Virginia and includes a small part of northernmost North Carolina. Sediments of the coastal plain underlie the eastern three-fifths of this area. These sediments onlap crystalline basement rocks toward the west and dip gently to the east, reaching a maximum known thickness of 821 feet in the extreme southeastern part of the map area. The gentle eastward dip is disrupted in several areas due to faulting delineated during the course of mapping. In order to produce a new geologic map of the Emporia 1:100,000-scale quadrangle, the U.S. Geological Survey drilled one corehole to a depth of 223 feet and augered 192 shallow research test holes (maximum depth 135 feet) to supplement sparse outcrop data available from the coastal plain part of the map area. The recovered sediments were studied and data from them recorded to determine the lithologic characteristics, spatial distribution, and temporal framework of the represented coastal plain stratigraphic units. These test holes were critical for accurately determining the distribution of major geologic units and the position of unit boundaries that will be shown on the forthcoming Emporia geologic map, but much of the detailed subsurface data cannot be shown readily through this map product. Therefore, the locations and detailed descriptions of the auger test holes and one corehole are provided in this open-file report for geologists, hydrologists, engineers, and community planners in need of a detailed shallow-subsurface stratigraphic framework for much of the Emporia map region.

  1. Fabrication characteristics of zircaloy tubes for nuclear reactors

    International Nuclear Information System (INIS)

    The production sequence for zircaloy cladding tubes to be used in nuclear reactors is described, with emphasis on the texture after reduction and on the variation in the hydrides orientation. The qualities requested for the cladding tubes are presented and reference is made to the quality control applied in the process. The destructive tests as well as the final inspection to which those tubes are subjected are related. A Fabrication Quality Project is requested from the manufacturers by reason of what Quality Control Plans are submitted to be clients. At last an evaluation of the quality to be obtained and of the control performed is mentioned. (Author)

  2. Baseline coastal oblique aerial photographs collected from Owls Head, Maine, to the Virginia/North Carolina border, May 19-22, 2009

    Science.gov (United States)

    Morgan, Karen L.M.; Hapke, Cheryl J.; Himmelstoss, Emily A.

    2015-01-01

    The U.S. Geological Survey (USGS) conducts baseline and storm response photography missions to document and understand the changes in vulnerability of the Nation's coasts to extreme storms. On May 19-22, 2009, the USGS conducted an oblique aerial photographic survey from Owls Head, Maine, to the Virginia/North Carolina border aboard a Cessna 207A at an altitude of 500 feet (ft) and approximately 1,200 ft offshore. This mission was flown to collect baseline data for assessing incremental changes since the last survey, and the data can be used in the assessment of future coastal change.

  3. Welding of tube-to-tube joints between martensitic and austenitic stainless steels for reactor applications

    International Nuclear Information System (INIS)

    Studies of the welding of tube-to-tube joints between dissimilar steels are described. The material combination was martensitic F6NM (0Cr13Ni4Mo) and austenitic AISI 347 stainless steels. Such joints are important parts in the primary circuit of a pressurized water reactor. A welding procedure was developed based on a series of weldability tests. The joints were subjected to various mechanical property tests including tensile test, bending test and impact test at various temperatures. A prototype product was made and subjected to pressure testing. Results of various tests indicated that the quality of the tube-to-tube joints is satisfactory for meeting all the design requirements. The joints can be used in the primary circuit of a pressurized water reactor. (author)

  4. When folic acid fails: Insights from 20 years of neural tube defect surveillance in South Carolina.

    Science.gov (United States)

    Bupp, Caleb P; Sarasua, Sara M; Dean, Jane H; Stevenson, Roger E

    2015-10-01

    Neural tube defects (NTDs) are the most common of the severe malformations of the brain and spinal cord. Increased maternal intake of folic acid (FA) during the periconceptional period is known to reduce NTD risk. Data from 1046 NTD cases in South Carolina were gathered over 20 years of surveillance. It was possible to determine maternal periconceptional FA use in 615 NTD-affected pregnancies. In 163 occurrent (26.9%) and two recurrent (22%) NTD cases, the mothers reported periconceptional FA use. These women were older and more likely to be white. Maternal periconceptional FA usage was reported in 40.4% of cases of spina bifida with other anomalies but in only 25.2% of isolated spina bifida cases (P = 0.02). This enrichment for associated anomalies was not noted among cases of anencephaly or of encephalocele. Among the 563 subsequent pregnancies to mothers with previous NTD-affected pregnancies, those taking FA had a 0.4% NTD recurrence rate, but the recurrence without FA was 8.5%. NTDs with other associated findings were less likely to be prevented by FA, suggesting there is a background NTD rate that cannot be further reduced by FA. Nonetheless, the majority (73.9%) of NTDs in pregnancies in which the mothers reported periconceptional FA use were isolated NTDs of usual types. Cases in which FA failed in prevention of NTDs provide potential areas for further study into the causation of NTDs. The measures and techniques implemented in South Carolina can serve as an effective and successful model for prevention of NTD occurrence and recurrence. PMID:26108864

  5. Remote dismantling of the pressure tube reactor from NPP Niederaichbach

    International Nuclear Information System (INIS)

    The pressure tube reactor of NPP Niederaichbach will be dismantled, segmented and packaged by remote operation, using a rotary manipulator, a cutting manipulator and a crane manipulator. With help from a number of remote controlled tools the rotary manipulator disassembles and lifts the reactor parts to a hot cell installed upon the upper reactor floor. Handling, crushing and packaging of those parts is performed with help from the crane manipulator. The cutting manipulator serves for segmenting of the moderator tank and the neutron shield tank

  6. Process and device for change of catalyst in tube reactors

    International Nuclear Information System (INIS)

    The change of catalyst in narrow reactor tubes with a height: diameter ratio of at least 30:1 is done by the catalyst filling being driven out against the force of gravity using a pulsating liquid flow. Pauses in the flow of between 0.1 to 1 sec between flow periods of 2 to 20 secs are useful. (orig./PW)

  7. Heavy-water-moderated pressure-tube reactor safety

    International Nuclear Information System (INIS)

    Several countries have heavy-water-moderated, pressure-tube reactors either in commercial operation or in late prototype stages. The supporting safety research and development includes such areas as the thermohydraulics of circuit depressurization, heat transfer from the fuel, heat rejection to the moderator from dry fuel, fuel and sheath behaviour, and fuel channel integrity. We review the work done in Canada, Great Britain, Italy and Japan, and describe some of the experimental tests underlaying the methods of accident analysis. The reactors have safety systems which, in the event of an accident, are able to shut down the reactor, keep the fuel cooled, and contain any released radioactivity. We summarize the characteristics of these safety systems (shutdown, emergency core cooling, and containment) in the various reactors, and discuss other reactor characteristics which either prevent accidents or reduce their potential demand on the safety systems. (author)

  8. Analysis of Reactor Vessel Lower Head Penetration Tube Failure

    International Nuclear Information System (INIS)

    This paper presents results of two studies, performed to investigate the behavior of the reactor vessel penetration tubes in case of relocation of molten material into the tubes. The first study is on the CORVIS drain line experiment 03/1. Results of pre-test calculations are presented, and compared to the later obtained experimental data. The timing of the drain line melting and the velocity of the debris flowing inside the drain line were predicted correctly, but the penetration depth was clearly underestimated. If the calculations are done using different correlation for the melt-to-wall convective heat transfer, the results are closer to the experiment. It cannot however be concluded that the alternative correlation is more appropriate until other uncertainties are clarified. The second study presents calculations performed for GKN Dodewaard CRD, instrument tubes and drain line. Calculations were performed to estimate whether the tubes have a chance to withstand the first attack of the melt and thus postpone vessel failure until the water in the lower plenum evaporates. Calculations were performed assuming that the melt can move into the tubes without any resistance, e.g. presence of water in the tubes was not taken into account. The results indicate that the critical penetration of the GKN vessel, which is most likely to fail, is the drain line. Results also indicate that external flooding should prevent early tube failure, at least in case of low vessel pressure. (author)

  9. Some design features of SCW pressure-tube nuclear reactor

    International Nuclear Information System (INIS)

    Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with supercritical water (SCW) became attractive again as the ultimate development path for water-cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern nuclear power plants (NPPs) from 33-35% to about 43-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (∼$1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625degC), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems. (author)

  10. Safety Evaluation Report related to renewal of the operating license for the CAVALIER Training Reactor at the University of Virginia (Docket No. 50-396)

    International Nuclear Information System (INIS)

    This Safety Evaluation Report for the application filed by the University of Virginia for a renewal of Operating License R-123 to continue to operate the CAVALIER (Cooperatively Assembled Virginia Low Intensity Educational Reactor) has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the University of Virginia and is located on the campus in Charlottesville, Virginia. Based on its technical review, the staff concludes that the reactor facility can continue to be operated by the university without endangering the health and safety of the public or the environment

  11. COMMODITY SCALE SYNTHESIS OF 1-METHYLIMIDAZOLE BASED IONIC LIQUIDS USING A SPINNING TUBE-IN-TUBE REACTOR

    Science.gov (United States)

    The continuous large-scale preparation of several 1-methylimidazole based ionic liquids was carried out using a Spinning Tube-in-Tube (STT) reactor (manufactured by Kreido Laboratories). This reactor, which embodies and facilitates the use of Green Chemistry principles and Proce...

  12. Education and training activities at North Carolina State University's PULSTAR reactor

    International Nuclear Information System (INIS)

    Research reactor utilization has been an integral part of the North Carolina State University's (NCSU's) nuclear engineering program since its inception. The undergraduate curriculum has a strong teaching laboratory component. Graduate classes use the reactor for selected demonstrations, experiments, and projects. The reactor is also used for commercial power reactor operator training programs, neutron radiography, neutron activation analysis (NAA), and sample and tracer activation for industrial short courses and services as part of the university's land grant mission. The PULSTAR reactor is a 1-MW pool-type reactor that uses 4% enriched UO2 pellet fuel in Zircaloy II cladding. Standard irradiation facilities include wet exposure ports, a graphite thermal column, and a pneumatic transfer system. In the near term, general facility upgrades include the installation of signal isolation and computer data acquisition and display functions to improve the teaching and research interface with the reactor. In the longer term, the authors foresee studies of new core designs and the development of beam experiment design tools. These would be used to study modifications that may be desired at the end of the current core life and to undertake the development of new research instruments

  13. Steam generator tube performance. Experience with water-cooled nuclear power reactors during 1985

    International Nuclear Information System (INIS)

    The performance of steam generator tubes at water-cooled reactors during 1985 has been reviewed. Seventy-three of 168 reactors in the survey experienced tube degradation sufficient for the tubes to be plugged. The number of tubes plugged was 6837 or 0.28% of those in service. The leading cause of tube failure was stress corrosion cracking from the primary side. Stress corrosion cracking or intergranular attack from the secondary side and pitting were also major causes of tube failure. Unlike most previous years, fretting was a substantial problem at some reactors. Overall, corrosion continued to account for more than 80% of the defects. 20 refs

  14. Wet channel measurement of pressure tube to calandria tube spacing in CANDU reactors

    International Nuclear Information System (INIS)

    The pressure tube (PT) to calandria tube (CT) spacing in CANDU reactors is an important parameter that relates to the general condition of the fuel channels. The measurement system that was developed to measure this parameter during the wet channel inspections of Pickering Units 1 and 2 is described in this paper. A send-receive eddy current probe was designed which is primarily sensitive to variations in PT/CT spacing but is also affected by pressure tube wall thickness. A computer simulation showed that the phase angles of the response to these variables are similar for all usable frequencies, thus eliminating the possibility of multifrequency compensation. A marriage of technologies was proposed involving the ultrasonic measurement of wall thickness values which are then used to extract the spacing information from the eddy current signal. The accuracy of the system is approximately ±(30% +.1mm) which has been sufficient to determine if and where any of the pressure tubes have come in contact with their calandria tube. Field experience with the new system is discussed and areas for development are also outlined

  15. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    International Nuclear Information System (INIS)

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D2O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D2O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

  16. Hydrodynamics of a self-agitated draft tube airlift reactor

    Directory of Open Access Journals (Sweden)

    Tekić Miodrag N.

    2014-01-01

    Full Text Available The main hydrodynamic characteristics of a novel-constructed, self-agitated draft tube airlift reactor (DT-ALR were investigated. Ten impellers, driven only by the means of gas throughput and induced liquid circulation, were inserted in the draft tube. The insertion of impellers caused bubble breakup and reduction of both mean bubble size and coalescence, even under the conditions of high gas throughputs. Although the impellers induced energy losses, the resistance to the flow was relatively lower due to their rotation, unlike the internals used in other research reported in the literature. In comparison to the conventional configuration of a DT-ALR, it was found that the presence of impellers led to significant changes in hydrodynamics: riser gas holdup and mixing time increased, while overall gas holdup and liquid velocity in the downcomer decreased. [Projekat Ministarstva nauke Republike Srbije, br. 172025

  17. Ultrasonic crack-tip diffraction in CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    Currently there is no reliable method of measuring defect depths in CANDU reactor pressure tubes. The demonstrated success of crack-tip diffraction (or time-of-flight-testing) in round-robins on thick components has promoted an interest in this technique. In CANDU reactors, pressure tubes are effectively accessible only from the inside. Development work has concentrated on outside surface defects using 45 degree shear waves in contrast to the longitudinal waves usually used for testing thick components with this technique. Due to the small wall thickness of the pressure tubes (4.2 mm) and the typical sizes of defects of interest (0.15 mm or greater), frequencies of the order of 20 MHz are being used. A further complication comes from the orientation of the defects, which may be at any angle in pressure tubes. Initial studies have been performed on a series of outside surface notches and slots, plus a real fatigue crack. This crack was on the inside surface, so the technique required measuring this defect's depth from the outside. Initial results are encouraging. Even without signal processing, crack-tip diffracted signals were detectable from all but very large (2.5 mm) and very small (less than 0.076 mm) notches. Errors in estimates of defect depths were typically less than 0.1 mm for all the notches, and the results were consistent. Measurements on the fatigue crack showed similar random errors, though there appeared to be a deterministic error of about 0.1 mm as well

  18. 78 FR 32442 - Record of Decision for the General Management Plan, Blue Ridge Parkway, Virginia and North Carolina

    Science.gov (United States)

    2013-05-30

    ... National Park Service Record of Decision for the General Management Plan, Blue Ridge Parkway, Virginia and... (NPS) announces the availability of the Record of Decision (ROD) for the General Management Plan (GMP... modern-day management realities. Under Alternative C, parkway management would be more integrated...

  19. Safety evaluation report related to the renewal of the operating license for the University of Virginia open-pool research reactor. Docket No. 50-062

    International Nuclear Information System (INIS)

    This Safety Evaluation Report for the application filed by the University of Virginia for a renewal of Operating Licence R-66 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned by the University of Virginia and is located on the campus in Charlottesville, Virginia. Based on its technical review, the staff concludes that the reactor facility can continue to be operated by the University without endangering the health and safety of the public or endangering the environment

  20. Towards a sustainable future using pressure tube reactor technology

    International Nuclear Information System (INIS)

    We describe the contribution nuclear energy will make to global energy needs based on the sound foundation of existing technology, infrastructure, natural resources and human knowledge, while meeting the requirements of security of supply (energy independence) and growing demand. Currently all reactors internationally operate on an unsustainable once-through nuclear fuel cycle using uranium fuel. Future decisions will be increasingly based on strategic considerations involving the complete nuclear fuel cycle, including requirements related to supply assurances, resource utilization, proliferation resistance and radioactive waste disposal. Pressure tube reactor (PTR) technology using fuel channels is uniquely suited to respond to the future needs because of its inherent technical characteristics and associated fuel cycle flexibility. PTR channel technology concepts have also continued to advance based on 50 years of continuous development and improvement, with strategic considerations involving the complete nuclear fuel cycle related to: Fuel Availability and Supply Assurances, Uranium, Plutonium and Thorium utilization, Waste Minimization, Proliferation Resistance (Safeguards) ,Assured Licensability, Improved Safety Cost, Competitiveness. We show how nuclear technology development and global sustainability is determined by R and D progress, with challenging technology goals for nuclear energy systems in the four areas of sustainability, economics, safety and reliability, and proliferation resistance and physical protection, leading naturally to the next phase of PTR channel development, namely the high efficiency Supercritical Water Reactor (SCWR). Aggressive targets have been set for R and D and advanced concepts, complementary to the approaches taken in India, which support enhanced safety, cost reduction, resource sustainability, and economical and efficient operation. (author)

  1. Contrasting styles of Hurricane Irene washover sedimentation on three east coast barrier islands: Cape Lookout, North Carolina; Assateague Island, Virginia; and Fire Island, New York

    Science.gov (United States)

    Williams, H. F. L.

    2015-02-01

    Storm surge and wind-driven waves generated by Hurricane Irene, which made landfall on the U.S. east coast on August 27 2011, resulted in overwash of sandy barrier islands from North Carolina to New York State. Overwash has significant impacts on barrier island geomorphology: it represents a sediment pathway into island interiors, a component of island sediment budgets, and can cause considerable aggradation of backshore surfaces, important for potentially offsetting the effects of rising sea level. This study describes the morphology, texture and microfossil content of Hurricane Irene washover deposits at three contrasting barrier island sites: Cape Lookout, North Carolina, Assateague Island, Virginia and Fire Island, New York. At all three sites, run-up overwash occurred, wherein waves were sufficient to overtop parts of the beach system and transport sediment inland. However, at Fire Island, overwash was restricted by a higher elevational threshold to low spots in the beach system coinciding with pre-existing breaches in foredunes. The result was the formation of isolated, thinner, low-volume washover fans. At Assateague Island and Cape Lookout, lower elevational thresholds allowed waves to overtop longer continuous sections of beach systems, resulting in the formation of laterally-continuous, thicker, larger-volume washover terraces. Overall, the deposits lacked consistent trends in thickness and texture (such as thinning and fining inland, reflecting a progressive reduction in overwash competence). Thickness and texture of the deposits were both spatially variable and probably reflect infilling of low points on the former surface and the influence of beach and foredune sediment sources. All the washover deposits were essentially barren of foraminiferal microfossils, supporting the textural evidence that the adjacent beach and foredunes were the predominant sediment sources.

  2. Formerly utilized MED/AEC sites remedial action program: radiological survey of the former Virginia-Carolina Chemical Corporation Uranium Recovery Pilot Plant, Nichols, Florida. Final report

    International Nuclear Information System (INIS)

    The results of a radiological survey conducted at the site of a former uranium recovery pilot plant operated by the Virginia-Carolina Chemical Corporation is presented. All that remains of this operation is a concrete pad situated within the boundary of a phosphate products plant now operated by Conserv, Inc., at the Nichols, Florida site. The survey included measurements designed to characterize the residual radioactivity in the vicinity of this pilot plant and to compare the quantities with federal guidelines for the release of decontaminated property for unrestricted use. The results of this survey indicate that only small quantities of radioactivity exist above normal background levels for that area. Some soil contamination was found in the vicinity of a concrete pad on which the pilot plant stood. Much of this contamination was due to 226Ra and 238U. Some beta-gamma dose rates in excess of applicable guidelines were observed in this same area. External gamma-ray exposure rates at 1 m above the ground range from 20 to 100 μR/hr. None of the direct measurements of alpha contamination were above guideline levels

  3. 137Cs dynamics within a reactor effluent stream in South Carolina

    International Nuclear Information System (INIS)

    Radiocesium dynamics are being studied in a blackwater creek which had received production reactor releases from the Savannah River Plant in South Carolina. Most 137Cs in the water column is dissolved or in colloidal form and is believed to originate primarily through outflow from an upstream ''contaminated'' reservoir. All ecosystem components in the stream have high 137Cs concentration factors. Radiocesium concentrations are highest in filamentous algae (332 pCi/g-dry) and suspended particulate matter (100 to 200 pCi/g). Other food chain bases had much lower 137Cs levels. Most consumer populations averaged 10 to 50 pCi/g. Radiocesium concentrations decreased in transfers between food chain bases and primary consumers or filter feeders. Omnivores and small predators have similar 137Cs concentrations with bioaccumulation occurring by top-carnivores. Radiocesium levels are around 100 pCi/g in largemouth bass and water snakes. Foodweb components in the stream have reached a dynamic equilibrium in 137Cs concentrations despite a 10 yr absence of reactor operations. Radiocesium levels are apparently being maintained through long-term 137Cs cycling in the upstream reservoir and surrounding flood plain forest systems. Rainfall and other physical processes influence the seasonal 137Cs fluctuations in stream components

  4. Irradiation of zircaloy calandria tube samples for growth measurement in Dhruva reactor

    International Nuclear Information System (INIS)

    In our PHWRs, the calandria tubes are made of seamwelded zircaloy-2 material which is placed in service in fully annealed condition. The available literature contains considerable information on the irradiation response of seamwelded calandria tubes thereby generating enough confidence towards assuming satisfactory behaviour of seamwelded calandria tubes in the reactor. However it is known that garter spring transmits a load from the pressure tube to the calandria tube and after a few years of reactor operation, creep of the calandria tube is the controlling factor in resisting the sag of the coolant channels. There is a proposal from Nuclear Fuel Complex (NFC), Hyderabad to manufacture the seamless zircaloy calandria tubes in place of welded tubes which will result in increased production of calandria tubes at a much lower cost and will also meet the increased production demand for calandria tubes for future power programme. In view of the necessity to change over from seamwelded to seamless zircaloy calandria tubes, it is very important to generate the irradiation creep, growth and mechanical properties data for seamless and seamwelded zircaloy calandria tubes by irradiating their samples in the reactor. One such irradiation programme has been started in Dhruva reactor where the samples of seamless and seamwelded zircaloy calandria tubes have been loaded in specially designed tiers in a slug rod and the slug rod has been installed in the reactor. Dhruva being a high flux reactor, the irradiation growth behaviour of seamless and seamwelded zircaloy sample can be studied in a much shorter time. This paper describes the details of the zircaloy sample irradiation programme including the design of special zircaloy sample tiers, the zircaloy slug rod and a precision zircaloy irradiation growth measuring jig towards collecting a meaningful zircaloy irradiation growth data. (author). 4 figs

  5. Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

    International Nuclear Information System (INIS)

    When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described

  6. Computation and measurement of calandria tube sag in pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Calandria tubes and liquid injection shutdown system nozzles in a pressurized heavy water reactor are to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with liquid injection shutdown system tube crossing beneath the calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measurement probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sag between both tubes in the reactor. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted

  7. Ultrasonic measurement of gap between calandria tube and liquid injection nozzle in CANDU reactor

    International Nuclear Information System (INIS)

    Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor(Calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the Calandria tube(made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, ti possibly contacts with liquid injection nozzle crossing beneath the Calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, Calandria tube and liquid injection tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here. The gap between LIN and CT at site was measured using by this ultrasonic device at site

  8. Non destructive examination of Reactor DR-3. Reactor wall, horisontal experimental tubes, up- and down comers

    International Nuclear Information System (INIS)

    The initial scope of work was to perform thickness/corrosion measurements of one up-comer and one down-comer, perform thickness/corrosion measurements in selected areas of the reactor wall and horizontal experimental pipes inside the reactor. Furthermore the lower circumferential weld and the connected longitudinal weld should be inspected to the extent possible, without major changes of the manipulator. Eddy current was performed in the same areas. Also hardness tests were carried out in four positions inside the reactor. Due to the outcome of the above examinations, additional metallurgical and dye penetrant examinations (PT) were carried out. The examination of the up- and down comers showed no sign of serious service induced defects. The eddy current testing did not reveal any inner surface breaking defects. The thickness/corrosion ultrasonic measurement showed only minor local indications with small or no reductions of original nominal wall thickness. The examination of the horizontal tubes showed no sign of serious service induced defects. The eddy current testing did not reveal any inner surface breaking defects. The thickness/corrosion ultrasonic measurement showed only minor local indications with small or no reductions of original nominal wall thickness. The hardness test showed increased hardness compared to calibration values. The examination of the reactor wall base material revealed several indications located in different depths in the plate. Some indications have been proved to be connected to the inner surface, while most indications appear to be either inclusions or areas corroded from the outside reactor wall. Minimum measured wall thickness is between 4.2 and 11.0 mm. There is, however, no evidence that these values are caused by corrosion at the outer reactor surface. The ET showed no signs of service induced cracks. The hardness test showed values close to calibration values. The extensive number of indications has resulted in additional

  9. CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

  10. The blister phenomena in relation to pressure tube integrity in CANDU reactor

    International Nuclear Information System (INIS)

    Zirconium alloy pressure tubes in CANDU type PHWR reactors are exposed to aqueous conditions embracing high temperature, fast neutron flux and high pressure. Two properties, dimension and hydrogen concentration, represent the main properties where changes are important to the life of a pressure tube. Rupture of a cold worked Zircaloy-2 pressure tube in Pickering Unit 2 in 1983 occurred when a crack developed from an array of hydride blisters. These have been observed on the outside surface of the pressure tube where it contacted the surrounding calandria tube. The contact of the pressure tube with the calandria tube can occur during the operation time and produces in the pressure tube a localized cooling. The consequence of the local heating of the calandria tube is some localized hydride precipitation. Under certain conditions, hydrogen will migrate down the temperature gradient and accumulate in the coldest region. Such precipitation, when it occurs under operating conditions, is considered to be the start of blister formation. When the blister cracking threshold is reached, the blister cracking can initiate a crack on the tube body. The failure mechanisms in zirconium alloy pressure tubes involve the presence of hydrogen in the initiation process and then a propagation process. If crack, originating from a hydride blister on the outside of the pressure tube, is developed then crack growth is possible in the axial direction to a partial thickness unstable length. Unstable pressure tube rupture is an event in a CANDU reactor that has potentially serious economic and safety consequences and reactor operation under conditions which entail the risk of such failure should be avoided. (authors)

  11. High-speed roll-to-roll manufacturing of graphene using a concentric tube CVD reactor

    OpenAIRE

    Polsen, Erik S.; McNerny, Daniel Q.; B. Viswanath; Pattinson, Sebastian W.; A. John Hart

    2015-01-01

    We present the design of a concentric tube (CT) reactor for roll-to-roll chemical vapor deposition (CVD) on flexible substrates, and its application to continuous production of graphene on copper foil. In the CTCVD reactor, the thin foil substrate is helically wrapped around the inner tube, and translates through the gap between the concentric tubes. We use a bench-scale prototype machine to synthesize graphene on copper substrates at translation speeds varying from 25 mm/min to 500 mm/min, a...

  12. Clinch River Breeder Reactor Plant Steam Generator Few Tube Test model post-test examination

    International Nuclear Information System (INIS)

    The Steam Generator Few Tube Test (FTT) was part of an extensive testing program carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. This paper describes the problems encountered with the mechanical features of the FTT model design which led to premature test termination, and the results of the post-test examination. Conditions of tube bowing and significant tube and tube support gouging was observed. An interpretation of the visual and metallurgical observations is also presented. The CRBRP steam generator has undergone design evaluations to resolve observed deficiences found in the FFTM

  13. Ultrasonic inspection of liquid-metal fast breeder reactor steam generator duplex tubing

    International Nuclear Information System (INIS)

    Two ultrasonic inspections of the Experimental Breeder Reactor II steam generator duplex tubing have been completed. Inspections performed on one evaporator in 1976 provided baseline data, and a subsequent inspection in 1978 revealed no change in tube condition. With the completion of the 1978 inspection, all available tubes in one evaporator have been inspected. The steam generator contains duplex tubes fabricated from 2 1/4 Cr-1 Mo ferritic steel. Access to the bore (water) side of the tubes was gained through the steam outlet piping. The inspection included a complete volumertic (100% of the tube material) examination, measurement of wall thickness, and evaluation of the condition of the braze bonding the two walls of the tube together. The test equipment was routinely calibrated against a standard containing artificial flaws. Artificial flaws as small as 1.6 mm long x 0.25 mm deep were readily detected

  14. IN-SITU MONITORING OF PRODUCT STREAMS FROM A SPINNING TUBE-IN-TUBE REACTOR USING A METTLER-TOLEDO REACT-IR

    Science.gov (United States)

    A Mettler-Toledo ReactIR system has been used for in-line, real-time monitoring of the product stream from a spinning tube-in-tube reactor (STT®, Kreido Laboratories, Camarillo California). This combination of a process intensified continuous-flow reactor and an in-situ analytic...

  15. Erythorbic acid promoted formation of CdS QDs in a tube-in-tube micro-channel reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liang, Yan; Tan, Jiawei; Wang, Jiexin; Chen, Jianfeng [State Key Laboratory of Organic–Inorganic Composites, Beijing University of Chemical Technology, Beijing 100029 (China); Research Center of the Ministry of Education for High Gravity Engineering and Technology, Beijing University of Chemical Technology, Beijing 100029 (China); Sun, Baochang, E-mail: sunbc@mail.buct.edu.cn [State Key Laboratory of Organic–Inorganic Composites, Beijing University of Chemical Technology, Beijing 100029 (China); Research Center of the Ministry of Education for High Gravity Engineering and Technology, Beijing University of Chemical Technology, Beijing 100029 (China); Shao, Lei, E-mail: shaol@mail.buct.edu.cn [State Key Laboratory of Organic–Inorganic Composites, Beijing University of Chemical Technology, Beijing 100029 (China); Research Center of the Ministry of Education for High Gravity Engineering and Technology, Beijing University of Chemical Technology, Beijing 100029 (China)

    2014-12-15

    Erythorbic acid assistant synthesis of CdS quantum dots (QDs) was conducted by homogeneous mixing of two continuous liquids in a high-throughput microporous tube-in-tube micro-channel reactor (MTMCR) at room temperature. The effects of the micropore size of the MTMCR, liquid flow rate, mixing time and reactant concentration on the size and size distribution of CdS QDs were investigated. It was found that the size and size distribution of CdS QDs could be tuned in the MTMCR. A combination of erythorbic acid promoted formation technique with the MTMCR may be a promising pathway for controllable mass production of QDs.

  16. Estimation of impact pressure due to rupture in beam-tube for research reactor

    International Nuclear Information System (INIS)

    Neutrons have been used for studies in material sciences of physics, chemistry, metals and alloys, ceramics, polymers, and biological sciences. This application leads to build up research reactor all over the world. JRTR (Jordan Research and Training Reactor) which plans to build up in Jordan is multipurpose research reactor which is developed entirely with domestic technology to overseas. Thermal power is 5MW upgradable 10MW. JRTR have four horizontal beam tubes, 3 ST(Standard) and 1NR (Neutron radiography). The beam tube's cavities are filled with helium, purged regularly to prevent a build-up of radioactive gases and moisture. They are highly reliable because they have no moving parts. The beam tube embedded part is aligned with its corresponding beam tube in the reflector. Objective of this study is to describe water hammer phenomenon in beam tube and determine an impact pressure charged in end film of beam tube for accomplishing nuclear safety function of research reactor while beam tube is ruptured due to some accident such as earthquake. The water hammer was experimentally and analytically studied by Lai, Saruba, Ballanco, and Watters

  17. In-situ inspection of grooves in reactor tube sheet using a remotely operated cast impression taking device

    International Nuclear Information System (INIS)

    Utmost importance is given to the in-service inspection of critical components of a reactor to ensure its reliable performance during the reactor operation. This paper describes a cast taking device using cold setting resin to take impression of the grooves being made in the tube sheet for sparger tube installation in pressurised heavy water reactor. (author)

  18. Clinch River Breeder Reactor Plant steam generator: FEW tube test model post test examination

    International Nuclear Information System (INIS)

    The Steam Generator Few Tube Test (FTT) is part of an extensive testing program being carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. The problems encountered with the mechanical features of the FTT model design which led to premature test termination and the results of the post-test examination are described

  19. Safety evaluation report related to the renewal of the operating license for the research reactor at North Carolina State University

    International Nuclear Information System (INIS)

    This safety evaluation report (SER) summarizes the findings of a safety review conducted by the staff of the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation (NRR). The staff conducted this review in response to a timely application filed by North Carolina State University (the licensee or NCSU) for a 20-year renewal of Facility Operating License R-120 to continue to operate the NCSU PULSTAR research reactor. The facility is located in the Burlington Engineering Laboratory complex on the NCSU campus in Raleigh, North Carolina. In its safety review, the staff considered information submitted by the licensee (including past operating history recorded in the licensee's annual reports to the NRC), as well as inspection reports prepared by NRC Region H personnel and first-hand observations. On the basis of this review, the staff concludes that NCSU can continue to operate the PULSTAR research reactor, in accordance with its application, without endangering the health and safety of the public. 16 refs., 31 figs., 7 tabs

  20. In-reactor deformation of a pilgered cold-worked Zr-2.5 wt% Nb pressure tube

    International Nuclear Information System (INIS)

    Zr-2.5 wt% Nb pressure tubes in CANDU reactors are cold-drawn to achieve the desired mechanical strength and tube dimensions. As part of an assessment of possible alternative fabrication processes, a pressure tube was cold-worked by pilgering and its dimensional stability monitored during service in the U-1 loop in the NRU test reactor. Strain rates have been determined for the pilgered tube from measured diameter and length changes after 22,060 h of reactor operation. These rates were compared with those for a cold-drawn tube also tested in NRU, and with the calculated behaviour of current power reactor pressure tubes extrapolated to the NRU test conditions. The deformation rates of the pilgered tube were found to be only marginally inferior to those of cold-drawn tubes

  1. Corrosion of steams generator tubes in pressurized water reactors

    International Nuclear Information System (INIS)

    The paper summarizes the various corrosion phenomena which have impaired the reliability of PWR steam generators, and reviews the remedies adopted against them: relaxation of residual stresses in the tubes, improved specifications for water chemistry, selection of new materials

  2. Mathematical simulation of the RBMK reactor pressure tubes ruptures during accidents: Computer code and verification

    International Nuclear Information System (INIS)

    The multiple rupture of the pressure tubes is the most dangerous accident of the channel reactors. There are about 2,000 channels in the RBMK. There exist two potential scenarios: (1) the case of accident when a group of channels becomes overheated; and (2) the case of accident with a rupture of one tube and shock loads on several adjacent channels. The described model considers the prediction technique for potential ruptures according to the first scenario. The probabilistic approach was applied due to existing of substantial scatter and uncertainties in parameters determining pressure tubes deformations and failure in accidents. It was founded on the randomization of the deterministic solution for pressure tube-graphite system deformation and rupture for varied values of chosen chance characters. The mathematical model for the deterministic solution considers the deformation of the system consisting of the pressure tube from the zirconium alloy containing 2.5% of niobium, graphite hard contact rings and graphite blocks. It was solved the common plane strain boundary task. Tube deformation includes three stages: tube deformation until the radial clearance between the tube and graphite disappears; tube deformation with metal flow into the vertical clearance in hard contact rings slits after disappearing of the radial clearance; deformation of the pressure tube-graphite system after closure of the radial clearance up to graphite failure. The mathematical model for the 1st scenario is described. The approach for code verification is also described

  3. The EL-4 reactor. Changing of a pressure tube on a test loop

    International Nuclear Information System (INIS)

    Right from the beginning of the EL-4 project, the research convected with the overall design of the reactor was guided by the various technical specifications resulting from a justifiable concern about the reliability. The external and internal tubes of each layer situated in the reactor block had in particular to be interchangeable. The research alone into the dismantling of the external tube, i.e in fact the pressure tube, justified a certain number of full-scale tests on a model. The tests carried out under relevant conditions on a non-irradiated structure made it possible to define a complete ranger of of positioning and un-positioning sequences at a distance for such a pressure tube. (authors)

  4. Tube sheet structural analysis of intermediate heat exchanger for fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    The Prototype Fast Breeder Reactor 'Monju' is the first power generating fast breeder reactor in Japan. We have been designing the components of the plant for manufacturing. Among these is the intermediate heat exchanger (IHX) which exchanges heat between primary and secondary sodium loop. The tube sheet of IHX (shell to ligament junction) is a difficult area from the view point of structural strength design under elevated temperature. To validate the structural integrity of tube sheet we performed the series of inelastic analysis and tube sheet thermal shock test using test pieces and half scale model of actual design. The results of inelastic analyses showed there is little progressive deformation around shell to ligament structural discontinuous junction. Furthermore, thermal shock tests showed no increase of an accumulative deformation. By these analyses and experiments, structural reliability of tube sheet could be shown. (author)

  5. Influence of Draft Tube Diameter on Operation Behaviorof Air Lift Loop Reactors

    Directory of Open Access Journals (Sweden)

    Laith S. S. Al-Kuffe

    2010-01-01

    Full Text Available The ratio of draft tube to reactor diameters is of decisive importance for the operation behavior of air lift loop reactors. The influence of draft tube geometry was investigated with respect to oxygen mass transfer and mixing time. The diameter ratio was varied between 0.33 and 0.80. The measurements were performed in two loop reactors with liquid capacities of 11.775 and 26.49 liters using aqueous with solutions of different coalescence behavior. The results show that there is no single diameter ratio which would produce most favorable conditions for the two process parameters. With respect to the more important requirements of aerobic cultures, i.e high oxygen mass transfer and efficient mixing, a diameter ratio between 0.5 and 0.6 is recommended. If high liquid velocities in the draft tube are required a ratio of 0.6 should be used.

  6. Waterproofed Photomultiplier Tube Assemblies for the Daya Bay Reactor Neutrino Experiment

    OpenAIRE

    Chow, Ken,; Cummings, John; Edwards, Emily; Edwards, William; Ely, Ry; Hoff, Matthew; Lebanowski, Logan; Li, Bo; Li, Piyi; Lin, Shih-Kai; Liu, Dawei; Liu, Jinchang; Luk, Kam-Biu; Miao, Jiayuan; Napolitano, Jim

    2015-01-01

    In the Daya Bay Reactor Neutrino Experiment 960 20-cm-diameter waterproof photomultiplier tubes are used to instrument three water pools as Cherenkov detectors for detecting cosmic-ray muons. Of these 960 photomultiplier tubes, 341 are recycled from the MACRO experiment. A systematic program was undertaken to refurbish them as waterproof assemblies. In the context of passing the water leakage check, a success rate better than 97% was achieved. Details of the design, fabrication, testing, oper...

  7. Automatic Inspection of Nuclear-Reactor Tubes During Production and Processing, Using Eddy-Current Methods

    International Nuclear Information System (INIS)

    The possibilities of automatic and semi-automatic inspection of tubes using eddy-current methods are described. The paper deals in particular with modem processes, compared to the use of other non-destructive methods. The essence of the paper is that the methods discussed are ideal for objective automatic inspection. Not only are the known methods described, but certain new methods and their application to the detection of flaws in reactor tubes are discussed. (author)

  8. Organic Synthesis in a Spinning Tube-in-Tube (STT¢) Reactor

    Science.gov (United States)

    Continuous-flow reactors have been designed to minimize and potentially overcome the limitations of heat and mass transfer that are encountered in chemical reactors and further experienced upon scale up of a reaction. With process intensification, optimization of the reaction i...

  9. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  10. A Development of Preliminary Evaluating Model of Crept Pressure Tube Diameter for CANDU Reactor (2)

    International Nuclear Information System (INIS)

    Pressure tube of CANDU reactor can be expanded toward both radial and axial directions due to irradiation under the high pressure and temperature condition. As the irradiation period increases, the radial expansion due to creep of the pressure tube increases. The radial expansion of the pressure tube comes out the reduction of the coolability and it results in the power deration. Although the radial expansion of the pressure tube directly affect the safety and economy of the currently operated CANDU reactor, there is no domestic evaluation model to predict the pressure tube diameter. Accordingly, it is necessary to develop the prediction model of the pressure tube diameter and the is the motivation of this study. The objectives of the current work is to develop the basic evaluation model of the pressure tube diameter for CANDU reactor especially for Wolsong NPP (Nuclear Power Plant). In order to develop the diameter evaluation model, measured data for total 86 channels were collected from Wolsong NPP 1, 2, 3 and 4 and analyzed. Based on the provided data, the operational conditions such as a temperature, pressure and neutron flux along the axial direction were derived. All data were analysed to derive the correlation between the pressure tube diameter and the other operation parameters. The evaluation model of pressure tube diameter was modeled by using the neural network algorithm. Neural network algorithm has been widely used to derive the non-linear relation between the input and output data. The developed neural network model was learned based on the data from Wolsong NPP 2, 3, and 4 and was tested by using data from Wolsong NPP 1. The current project will be carried out by IAEA CRP in which all CANDU nations are going to participate

  11. Waterproofed photomultiplier tube assemblies for the Daya Bay reactor neutrino experiment

    Science.gov (United States)

    Chow, Ken; Cummings, John; Edwards, Emily; Edwards, William; Ely, Ry; Hoff, Matthew; Lebanowski, Logan; Li, Bo; Li, Piyi; Lin, Shih-Kai; Liu, Dawei; Liu, Jinchang; Luk, Kam-Biu; Miao, Jiayuan; Napolitano, Jim; Ochoa-Ricoux, Juan Pedro; Peng, Jen-Chieh; Qi, Ming; Steiner, Herbert; Stoler, Paul; Stuart, Mary; Wang, Lingyu; Yang, Changgen; Zhong, Weili

    2015-09-01

    In the Daya Bay Reactor Neutrino Experiment 960 20-cm-diameter waterproof photomultiplier tubes are used to instrument three water pools as Cherenkov detectors for detecting cosmic-ray muons. Of these 960 photomultiplier tubes, 341 are recycled from the MACRO experiment. A systematic program was undertaken to refurbish them as waterproof assemblies. In the context of passing the water leakage check, a success rate better than 97% was achieved. Details of the design, fabrication, testing, operation, and performance of these waterproofed photomultiplier-tube assemblies are presented.

  12. Remote metallurgical investigations on pressure tubes removed from CANDU power reactors

    International Nuclear Information System (INIS)

    As part of the periodic in-service inspection program for CANDU reactors, pressure tubes are periodically removed for destructive examination. The procedures, equipment, and facilities used to perform metallurgical examinations on these highly irradiated components are described. The initial examinations of the tubes from the generating station are performed underwater in inspection bays. Detailed visual examination and metallography are subsequently performed in shielded hot-cell facilities; a description of the remote metallographic equipment and preparation techniques used is given. Examinations of two recently removed Zr-2.5%Nb pressure tubes containing fretting-wear flaws and a lamination flaw are used to highlight the techniques employed

  13. Waterproofed photomultiplier tube assemblies for the Daya Bay reactor neutrino experiment

    International Nuclear Information System (INIS)

    In the Daya Bay Reactor Neutrino Experiment 960 20-cm-diameter waterproof photomultiplier tubes are used to instrument three water pools as Cherenkov detectors for detecting cosmic-ray muons. Of these 960 photomultiplier tubes, 341 are recycled from the MACRO experiment. A systematic program was undertaken to refurbish them as waterproof assemblies. In the context of passing the water leakage check, a success rate better than 97% was achieved. Details of the design, fabrication, testing, operation, and performance of these waterproofed photomultiplier-tube assemblies are presented

  14. Waterproofed photomultiplier tube assemblies for the Daya Bay reactor neutrino experiment

    Energy Technology Data Exchange (ETDEWEB)

    Chow, Ken [Engineering Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Cummings, John [Department of Physics and Astronomy, Siena College, Loudonville, NY 12211 (United States); Edwards, Emily [Engineering Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Edwards, William [Physics Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Ely, Ry [Department of Physics, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Hoff, Matthew [Engineering Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Lebanowski, Logan [Department of Physics, University of Houston, Houston, TX 77204-5005 (United States); Li, Bo; Li, Piyi [School of Physics, Shandong University, Jinan 250100 (China); Lin, Shih-Kai [Department of Physics, University of Houston, Houston, TX 77204-5005 (United States); Liu, Dawei [Department of Physics, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Liu, Jinchang [Key Laboratory of Particle Astrophysics, Institute of High Energy Physics, Beijing 100049 (China); Luk, Kam-Biu, E-mail: k_luk@berkeley.edu [Department of Physics, University of California, Berkeley, CA 94720 (United States); Physics Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Miao, Jiayuan [School of Physics, Shandong University, Jinan 250100 (China); Napolitano, Jim [Department of Physics, Temple University, Philadelphia, PA 19122 (United States); Ochoa-Ricoux, Juan Pedro [Physics Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Peng, Jen-Chieh [Department of Physics, University of Illinois at Urbana-Champaign, Urbana, IL 61801 (United States); Qi, Ming [Department of Physics, Nanjing University, Nanjing 210000 (China); and others

    2015-09-11

    In the Daya Bay Reactor Neutrino Experiment 960 20-cm-diameter waterproof photomultiplier tubes are used to instrument three water pools as Cherenkov detectors for detecting cosmic-ray muons. Of these 960 photomultiplier tubes, 341 are recycled from the MACRO experiment. A systematic program was undertaken to refurbish them as waterproof assemblies. In the context of passing the water leakage check, a success rate better than 97% was achieved. Details of the design, fabrication, testing, operation, and performance of these waterproofed photomultiplier-tube assemblies are presented.

  15. Waterproofed Photomultiplier Tube Assemblies for the Daya Bay Reactor Neutrino Experiment

    CERN Document Server

    Chow, Ken; Edwards, Emily; Edwards, William; Ely, Ry; Hoff, Matthew; Lebanowski, Logan; Li, Bo; Li, Piyi; Lin, Shih-Kai; Liu, Dawei; Liu, Jinchang; Luk, Kam-Biu; Miao, Jiayuan; Napolitano, Jim; Ochoa-Ricoux, Juan Pedro; Peng, Jen-Chieh; Qi, Ming; Steiner, Herbert; Stoler, Paul; Stuart, Mary; Wang, Lingyu; Yang, Changgen; Zhong, Weili

    2015-01-01

    In the Daya Bay Reactor Neutrino Experiment 960 20-cm-diameter waterproof photomultiplier tubes are used to instrument three water pools as Cherenkov detectors for detecting cosmic-ray muons. Of these 960 photomultiplier tubes, 341 are recycled from the MACRO experiment. A systematic program was undertaken to refurbish them as waterproof assemblies. In the context of passing the water leakage check, a success rate better than 97% was achieved. Details of the design, fabrication, testing, operation, and performance of these waterproofed photomultiplier-tube assemblies are presented.

  16. Effect of processing on properties of thin walled calandria tubes for pressurised heavy water reactor

    Science.gov (United States)

    Kapoor, K.; Padmaprabu, C.; Ramana Rao, S. V.; Sanyal, T.; Kashyap, B. P.

    2003-02-01

    Thin walled calandria tubes for pressurised heavy water reactors are manufactured either by seam welding of Zircaloy-4 sheets or by seamless route. In the present study, the effect of processing on the critical properties such as texture, microstructure, hydriding behaviour and residual stress for both the routes as well as the mechanical anisotropy developed due to seam welding are investigated. The properties of the seam welded tube in the fusion and adjoining region are markedly different from the base material and from the seamless tube. Residual stress measurements indicate that heat affected zone (HAZ) of seam welded tubes have longitudinal tensile residual stress and the seamless tubes have uniform compressive stress along the circumference. The phase transition in the presence of residual stresses due to thermal gradient is found to modify the texture in the HAZ. The hydride orientation and mechanical anisotropy in these regions are found to be dependent on the texture of the material.

  17. Effect of processing on properties of thin walled calandria tubes for pressurised heavy water reactor

    International Nuclear Information System (INIS)

    Thin walled calandria tubes for pressurised heavy water reactors are manufactured either by seam welding of Zircaloy-4 sheets or by seamless route. In the present study, the effect of processing on the critical properties such as texture, microstructure, hydriding behaviour and residual stress for both the routes as well as the mechanical anisotropy developed due to seam welding are investigated. The properties of the seam welded tube in the fusion and adjoining region are markedly different from the base material and from the seamless tube. Residual stress measurements indicate that heat affected zone (HAZ) of seam welded tubes have longitudinal tensile residual stress and the seamless tubes have uniform compressive stress along the circumference. The phase transition in the presence of residual stresses due to thermal gradient is found to modify the texture in the HAZ. The hydride orientation and mechanical anisotropy in these regions are found to be dependent on the texture of the material

  18. Calculation of fuel element temperature TRIGA 2000 reactor in sipping test tubes using CFD

    International Nuclear Information System (INIS)

    It has been calculated the fuel element temperature in the sipping test of Bandung TRIGA 2000 reactor. The calculation needs to be done to ascertain that the fuel element temperatures are below or at the limit of the allowable temperature fuel elements during reactor operation. ensuring that the implementation of the test by using this device, the temperature is still within safety limits. The calculation is done by making a model sipping test tubes containing a fuel element surrounded by 9 fuel elements. according to the position sipping test tubes in the reactor core. by using Gambit. Dimensional model adapted to the dimensions of the tube and the fuel element in the reactor core of Bandung TRIGA 2000 reactor. Sipping test Operation for each fuel element performed for 30 minutes at 300 kW power. Calculations were performed using CFD software and as input adjusted parameters of TRIGA 2000 reactor. Simulations carried out on the operation of the 30, 60, 90, 120, 150, 180 and 210 minutes. The calculation result shows that the temperature of the fuel in tubes sipping test of 236.06 °C, while the temperature of the wall is 87.58 °C. The maximum temperature in the fuel center of TRIGA 2000 reactor in normal operation is 650 °C. and the boiling is not allowed in the reactor. So it can be concluded that the operation of the sipping test device are is very safe because the fuel center temperature is below the temperature limits the allowable fuel under normal operating conditions as well as the fuel element wall temperature is below the boiling temperature of water. (author)

  19. Thickness measurement of A-1 reactor caisson tube walls

    International Nuclear Information System (INIS)

    The equipment is described of measuring the thickness of caisson pipes built in the Bohunice A-1 reactor. The pulse-type ultrasonic thickness gauge is based on the reflection method using the double probe. The measurement accuracy is 0.1 mm. (J.B.)

  20. Evaluation and analysis of critical crack length of irradiated pressure tubes from Indian pressurised heavy water reactors

    International Nuclear Information System (INIS)

    Results of fracture toughness KJic were computed from transverse tensile properties of reactor operated pressure tubes, and axial critical crack length values derived from KJic are presented. Similarly fracture resistance curves derived from tensile properties of reactor operated pressure tubes and axial critical crack length values computed therefrom are presented. Under normal operating condition of the reactors the pressure tubes experience temperatures ranging from 250 deg C to 300 deg C. In occurrence of contact between pressure tube and calandria tube, the contact region may not be expected to have a mean through wall temperature below 200 deg C. The axial critical crack length of three reactor operated pressure tubes, therefore were evaluated in the temperature range 200 deg C to 300 deg C. The significance of the magnitude of the evaluated critical crack length is discussed. (author)

  1. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  2. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis

  3. Characterisation of hydride blister in reactor operated zircaloy-2 pressure tube

    International Nuclear Information System (INIS)

    Zircaloy-2 pressure tubes pickup Hydrogen species (H and D) during in-reactor service. The hydrogen pickup leads to hydride precipitation and in the event of a contact between the pressure tube and the calandria tube, hydrogen migrates to the cold spot leading to the formation of hydride blister. One such hydride blister location in an operated Zircaloy-2 pressure tube of RAPS-2 was subjected to metallographic studies and mapping of the microstructure across the tube thickness. Mapping of the hydrogen concentration across the tube thickness was carried out by careful sampling and H estimation by DSC technique. The H profile across the tube thickness, up to the blister boundary, was generated. The hydride blister region was found to be made up of microstructurally different regions starting from dense massive hydride at the outer surface and followed in sequence by a region with dense and thick platelets oriented parallel to the blister boundary and radial platelet region, which subsequently merged with the background platelet distribution appropriate for the average hydrogen content of the pressure tube. The equivalent blister depth corresponding to H content of 16,000 w/ppm has been estimated from the H profile at the blister location. In the case of a hydride blister with measured thickness of 0.4mm the equivalent blister thickness was found to be 0.414mm. Mapping of the hardness of the massive hydride and the adjoining microstructurally different regions was carried out by microhardness measurements at room temperature. (author)

  4. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  5. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

  6. Ultrasonic systems for examining butt-fillet tube to tube plate welds in the evaporators of the Dounreay Prototype Fast Reactor

    International Nuclear Information System (INIS)

    Ultrasonic techniques and comprehensive data recording systems are described which have been developed for the in-service inspection of the tube to tube plate welds in the evaporator steam generator units of the Prototype Fast Reactor at Dounreay, Caithness, Scotland

  7. Measurement of internal diameter of pressure tubes in pressurized heavy water reactors using ultrasonics

    International Nuclear Information System (INIS)

    The Pressure Tube in Pressurized Heavy Water Reactors (PHWRs) undergoes dimensional changes due to the effects of creep and growth as it is subjected to high pressure and temperature, which causes Pressure Tubes to permanently increase in length and diameter and to sag because of weight of fuel and coolant (heavy water) contained in it. These dimensional changes are due to prolonged stresses under high temperature and radiation. Pressure Tube stresses are evaluated for both beginning and end of life for accounting the Pressure Tube dimensional changes that occur during its design life. At the beginning of life, the initial wall thickness and un-irradiated material properties are applied. At the end of life, Pressure Tube diameter and length increases, while wall thickness decreases. Material strength also increases during that period. The increase in Pressure Tube diameter results in squeezing of garter spring spacer between the pressure and calandria Tubes. It also causes unacceptable heat removal from the fuel due to an increased amount of primary coolant that bypasses the fuel bundles. This reduces the critical channel power at constant flow. Hence the periodic monitoring of pressure Tube diameter is important for these reasons. This is also required as per the applicable codes and standards for In-Service Inspection of PHWRs. Mechanical measurement from ID of the Tube during periodic monitoring is not practically feasible due to high radiation and inaccessibility. This necessitates the development of NDT technique using Ultrasonics for periodic in-situ measurement of ID of pressure Tubes with a BARC made remotely operated drive system called BARCIS (BARC Channel Inspection system). The development of Ultrasonic based ID measurement techniques and their actual applications in PHWRs Pressure tubes are being discussed in this paper. (author)

  8. A Feasibility Study of Surveillance of Pressure Tubes in Heavy Water Reactors

    International Nuclear Information System (INIS)

    All the Zr-2.5Nb pressure tubes used in all CANDU 6 plants and Wolsong nuclear power plants were made according to AECL's design to have a strong tangential texture for high creep resistance. Nevertheless, they are to be replaced before reaching their design lifetime due to higher growth and diametral creep than expected, the latter of which causes a reduction of the plant power to below the 100% full power after 15 years of operation. These instances of operational experience show that AECL's design of pressure tubes is invalid. Besides, comparison of in-reactor creep between pressure tubes with radial and tangential textures was made not on the condition of all variables being constant except texture but with both texture and the microstructure being varied. In other words, the in-reactor creep tests carried out by AECL turns out to be ineffective to single out the effect of texture on creep. In contract, comparison of in-reactor creep between CANDU Zr-2.5Nb tube and Russian Zr-2.5Nb tubes demonstrate that creep of Zr-2.5Nb tubes are governed not by texture but by another factor. Given that Russian TMT-2 tube with the lower degree of tangential texture shows two times lower creep rate than CANDU Zr-2.5Nb tube, it is suggested that the stability of Nb dissolved in the α-Zr grains is a creep controlling factor. Fortunately, given our improved pressure tubes being developed in cooperation with Russia that were made similar to the TMT-2 tube's manufacturing process but optimized for a better control of fine precipitates and texture, it is evident that our improved tubes will be better due to excellent creep resistance, higher fracture toughness and zero axial growth when compared to TMT-2 tube. Since the 2005 version of the CSA N285.4 stipulates surveillance of pressure tubes, material examination of Wolsong Unit-2 pressure tubes should be conducted since 2013. Korea Hydro Nuclear Power (KHNP)'s strategy is to prepare an alternative instead of material examination

  9. Underwater plasma arc cutting of in-reactor tube of In-Pile Creep Test Facility

    International Nuclear Information System (INIS)

    The in-reactor tube of the In-Pile Creep Facility had been irradiated periodically for over 6 years in the Japan Materials Testing Reactor (JMTR) up to the end of 1978 under an operating condition of high temperature and high pressure identical to that of the Prototype Advanced Thermal Reactor, FUGEN, to gain the basic data for estimating the amount of creep which would occur in the pressure tubes of FUGEN. Following the removal of the in-reactor tube out of the JMTR, the test sections in the tube which were to be subjected to post irradiation examination were cut out. Underwater plasma arc cutting was employed to prevent the spread of contamination to the work area, to confine the heat affected zone in the test pieces to a minimum and to simplify disposal of the unneeded portions of the pressure tube. Setup of the cutting machine, cutting operations, radiological conditions during cutting of the highly radioactive portion of the tube and disassembly of the cutting equipment are described. In addition a brief description of the underwater plasma arc cutting machine is presented. The hot-cutting operations were done remotely to control personal exposure. The containment envelope prevented the spread of contamination to the environment and radioactive particles deposited on the cutting machine were removed without any difficulties. External exposure received by cutting personnel was small. Internal radionuclide deposit examinations were conducted, determining no crew member inhaled radioactive substances. Contamination spreads to the work area were minimal and release of radionuclide was well controlled. (author)

  10. Testing of a 7-tube palladium membrane reactor for potential use in TEP

    International Nuclear Information System (INIS)

    A Palladium Membrane Reactor (PMR) consists of a palladium/silver membrane permeator filled with catalyst (catalyst may be inside or outside the membrane tubes). The PMR is designed to recover tritium from the methane, water, and other impurities present in fusion reactor effluent. A key feature of a PMR is that the total hydrogen isotope content of a stream is significantly reduced as (1) methane-steam reforming and/or water-gas shift reactions proceed on the catalyst bed and (2) hydrogen isotopes are removed via permeation through the membrane. With a PMR design matched to processing requirements, nearly complete hydrogen isotope removals can be achieved. A 3-tube PMR study was recently completed. From the results presented in this study, it was possible to conclude that a PMR is appropriate for TEP, perforated metal tube protectors function well, platinum on aluminum (PtA) catalyst performs the best, conditioning with air is probably required to properly condition the Pd/Ag tubes, and that CO/CO2 ratios maybe an indicator of coking. The 3-tube PMR had a permeator membrane area of 0.0247 m2 and a catalyst volume to membrane area ratio of 4.63 cc/cm2 (with the catalyst on the outside of the membrane tubes and the catalyst only covering the membrane tube length). A PMR for TEP will require a larger membrane area (perhaps 0.35 m2). With this in mind, an intermediate sized PMR was constructed. This PMR has 7 permeator tubes and a total membrane area of 0.0851 m2. The catalyst volume to membrane area ratio for the 7-tube PMR was 5.18 cc/cm2. The total membrane area of the 7-tube PMR (0.0851 m2) is 3.45 times larger than total membrane area of the 3-tube PMR (0.0247 m2). The following objectives were identified for the 7-tube PMR tests: (1) Refine test measurements, especially humidity and flow; (2) Refine maintenance procedures for Pd/Ag tube conditioning; (3) Evaluate baseline PMR operating conditions; (4) Determine PMR scaling method; (5) Evaluate PMR with realistic

  11. Process modifications in the manufacture of zirconium alloy clad tubes for nuclear reactors

    International Nuclear Information System (INIS)

    Full text: Indian Nuclear Power Reactors use fuel bundles containing Uranium di- oxide pellets clad in Zirconium alloy tubes. The fuel tubes required for these fuel bundles are subjected to severe irradiation and corrosive environment in the nuclear reactor in addition to the exposure of high temperatures. Hence they are manufactured conforming to strict specifications with respect to mechanical, chemical, metallurgical properties and close dimensional tolerances. The fuel tubes are normally manufactured by processing the extruded blanks using 3 or 4 stage cold pilgering with intermediate heat treatment and final finishing operations. This presentation gives a brief detail of process modifications made in the manufacture of these tubes, to improve the mechanical and metallurgical properties and also to improve the overall material recovery. It was established that the variation in wall thickness in the extruded blanks is carried forward to subsequent passes which affects the mechanical properties of the tubes. The annealing parameters were optimized to achieve finer grain size, resulting in better mechanical properties Some of the important modifications carried out include the following: 1. Reducing the wall thickness variation and improving the surface finish of extruded blanks by machining, 2. Introduction of ultrasonic testing of blanks, 3. Optimization of parameters at all stages of pilgering i.e., blank, intermediate and final pass stages, 4. Optimization of parameters at annealing and sand blasting operations, and 5. Improvisation of finishing operations

  12. A plan to preserve the environmental integrity of the False Cape Outer Banks area : Virginia and North Carolina [Back Bay National Wildlife Refuge

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — This document is a plan meant to gain control of certain lands in the False Cape area of Virginia. Various methods of acquiring the land and their pros and cons are...

  13. Investigation of surface corrosion layers of fast reactor heat exchanger tubes

    International Nuclear Information System (INIS)

    The potential reasons of fast reactor heat exchanger tubes destruction and the ways of corrosion inhibition are studied. Using the methods of Auger spectroscopy and X-ray diffraction microanalysis element distribution in depth of corrosion layers from coolant (sodium) side and from surface contacting with steam is investigated. It is shown that sodium is present through all thickness of the tube. Pulsed plasma treatment of steel 12Cr18Ni9 surface decreases intercrystalline corrosion susceptibility due to structural changes of surface layer (near 20 μm), its enrichment by chromium and protecting chromium oxide film formation

  14. Assessment of aging of Zr-2.5Nb pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    In modern CANDU nuclear generating stations, pressure tubes of cold-worked Zr-2.5Nb material are used in the reactor core to contain the fuel bundles and the heavy water (D2O) coolant. The pressure tubes operate at an internal pressure of about 10 MPa and temperatures ranging from about 250oC at the inlet to about 310oC at the outlet. Over the expected 30 year lifetime of these tubes they will be subjected to a total fluence of approximately 3 x 1026 n m-2. In addition, these tubes gradually pick up deuterium as a result of a slow corrosion process. When the hydrogen plus deuterium concentration in the tubes exceeds the hydrogen-deuterium solvus, the tubes are susceptible to a crack initiation and propagation process called delayed hydride cracking (DHC). If undetected, such a cracking mechanism could lead to unstable rupture of the pressure tube. A fitness-for-service methodology has been developed which assures that this will not happen. A key element in this methodology is the acquisition of data and understanding-from surveillance and accelerated aging testing-to assess and predict changes in the DHC initiation threshold, the DHC velocity and the fracture toughness (critical crack length) as a function of service time. The most recent results of the DHC and fracture toughness properties of CANDU pressure tubes as a function of time in service are presented and used to suggest procedures for mitigation and life extension of the pressure tubes. (author)

  15. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    International Nuclear Information System (INIS)

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures

  16. SOLAR REFRIGERATING UNIT WITH AN ADSORPTION REACTOR AND EVACUATED TUBE COLLECTORS

    Directory of Open Access Journals (Sweden)

    M.E. Vieira

    1997-09-01

    Full Text Available This work presents the principles of operation of a solar refrigerator with the following basic components: a reactor, a set of evacuated tube solar collectors, a condenser, a heat exchanger, and an evaporator. During the heating phase, solar radiation is collected and transferred to the reactor for desorption by a vapor thermal siphon loop. During the cooling phase, heat from the reactor is released to the ambient by a second water vapor loop. Ambient data collected daily during a period of 18 years were divided into hourly values and used to simulate the temperatures of the reactor, which uses salt impregnated with graphite and ammonia, during the adsorption / desorption processes. The results show that the refrigerator operates well in Fortaleza and that better results are expected for the countryside of the state of Ceara. It is concluded that only a high efficiency collector set can be used in the system

  17. Design of an ion transport membrane reactor for application in fire tube boilers

    International Nuclear Information System (INIS)

    A design of an ITM (ion transport membranes) reactor is introduced in a two-pass fire tube boiler furnace to produce steam for power generation toward the ZEPP (zero emission power plant) applications. Oxygen separation, combustion and heat exchange occur in the first pass containing the multiple-units ITM reactor. In the second pass, heat exchange between the combustion gases and the surrounding water at 485 K (Psat = 20 bar) occurs mainly by convection. The emphasis is to extract sufficient oxygen for combustion while maintaining the reactor size as compact as possible. Based on a required power in the range of 5–8 MWe, the fuel and gases flow rates were calculated. Accordingly, the channel width was determined to maximize oxygen permeation flux and keep the viscous pressure drop within a safe range for fixed reactor length of 1.8 m. Three-dimensional simulations were conducted for both counter and co-current flow configurations. Counter-current flow configuration proved its suitability in fire tube boilers for steam generation over the co-current flow configuration. The resultant reactor consists of 12,500 ITM units with a height of 5 m, membrane surface area of 2700 m2 and a total volume of 45.45 m3. - Highlights: • A novel two-path fire tube boiler design is presented utilizing ITMs (ion transport membranes). • A new multi-unit ITM reactor design for boiler furnace substitution is presented. • Flow rates have been optimized for maximum oxygen flux and power generation. • Counter-current flow configuration is much more efficient than co-current flow. • Total number of ITM units was calculated to produce power of 5:8 MWe

  18. Numerical analysis of zirconium hydride blisters in CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    CANDU nuclear reactors use zirconium alloy pressure tubes for primary containment of fuel and coolant. The 1983 failure of a pressure tube in Unit 2 of the Pickering Nuclear Generating Station was attributed to the formation of large precipitates of zirconium hydride, referred to as blisters. These blisters formed at localized cold spots on the pressure tube surface where it had come into contact with the colder calandria tube. The high hydrogen concentrations in the Zircaloy-2 pressure tubes used only in the first two Pickering Units were a major contributing factor to blister formation and the ultimate failure. In an effort to better understand the mechanism of crack initiation at a blister, a program was undertaken to use finite element methods to model the stresses generated by the formation of a blister in a tube. The preliminary results in this work have been published elsewhere. This paper summarizes the recent refinements to the model and our present understanding of the development of stresses in and around hydride blisters. (orig./GL)

  19. GERDA test facility for pressurized water reactors with straight tube steam generators

    International Nuclear Information System (INIS)

    A number of large-scale experimental facilities have been constructed and operate in order to experiment on the thermodynamic and thermohydraulic behaviour of nuclear facilities in case of LOCA. Most of them were designed for ''large leak'' accidents, but as ''small leak'' accidents became the focus of interest, such experiments were also carried out. Experiments carried out with this arrangement for PWR-type reactors with straight-tube steam generators are only partially evaluable. BBR and B and W therefore cooperated in the construction of the test facility GERDA, designed for testing reactors of BBR design. It supplied relevant experimental results for the nuclear power plant at Muelheim-Kaerlich. (orig.)

  20. Some aspects of the thorium fuel cycle in heavy-water-moderated pressure tube reactors

    International Nuclear Information System (INIS)

    The use of thorium fuel cycles in heavy-water-moderated pressure tube (CANDU) reactors will allow much more energy to be extracted from a given amount of fuel than is possible with the present natural uranium cycle. The extent to which various factors affect thorium fuel cycle economics and resource consumption with equilibrium 233U levels in the fuel is considered. Resource consumption in growing nuclear power systems is also considered, and it is shown that considerable savings can be achieved even under conditions of rapid growth. The main elements of the development program necessary to provide the technological base for thorium fuel cycles in CANDU reactors are discussed. (author)

  1. Some aspects of the thorium fuel cycle in heavy-water-moderated pressure tube reactors

    International Nuclear Information System (INIS)

    The use of thorium fuel cycles in heavy-water-moderated pressure tube (CANDU) reactors will allow much more energy to be extracted from a given amount of fuel than is possible with the present natural uranium cycle. The extent to which various factors affect thorium fuel cycle economics and resource consumption with equilibrium 233U levels in the fuel is considered. Resource consumption in growing nuclear power systems is also considered, and it is shown that considerable savings can be achieved even under conditions of rapid growth. The main elements of the development program necessary to provide the technological base for thorium fuel cycles in CANDU reactors are discussed

  2. Beam tube experiments and correlated research projects at the TRIGA reactor Vienna

    International Nuclear Information System (INIS)

    The four beam tubes and the thermal column at the TRIGA reactor Vienna were used intensively during the reporting period. Three of the beam tubes are mainly used for neutron spectroscopy such as small angle scattering, neutron interferometry and polarized neutrons where now investigations of magnetic structures in pulsed high magnetic fields (25 T) synchronized with the pulsed mode of the reactor have been started. The thermalizing column will be modified from the present cold neutron source to a comfortable neutron radiography installation which allows investigations of objects of a size up to 30 cm diameters. The thermal column is also used for neutron radiography and as a strong gamma source to investigate gamma irradiation effects on various materials such as glass fiber cables. In view of flexible utilization of the thermal column a movable shielding construction has been designed which is simple rolled away on the rails of the thermal column doors when access to the thermal column in necessary. (orig.)

  3. Study of neutron beam silhouette at tangential-through-tube of Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Pakistan Research Reactor-1 (PARR-1) provides facilities to conduct experiments of vital importance using thermal neutron beams derived from the reactor core. One tangential-through-tube and several neutron beam tubes are available around the reactor for researchers. At the tangential-through-tube of PARR-1, experimental facilities for Prompt Gamma Neutron Activities Analysis (PGNAA) have been indigenously established. While designing the collimator it was imperative to ensure a proper collimation of thermal neutron beam on the target being exposed. It was, therefore, required to observe the neutron beam silhouette at various sections of the tangential-through-tube. In a series of experiments, CR-39 track detectors were exposed with neutrons at various sections of tangential-through-tube for about an hour while thermal neutron flux was measured in the range (1.8 x 10 /sup 7/ to 3.2 x 10 /sup 8/ neutrons cm/sup -2/.s/sup -1/. Optimum etching conditions were experimentally obtained to provide the best neutron beam profiles. The icon of the beam silhouette on the detectors can easily be observed with the naked eyes. However, innovative attempts have been made to reproduce the neutron silhouette onto paper by scanning these detectors using Laser jet scanner-4. This paper displays several scanned photographs of thermal neutron beam silhouette. In one of the neutron beam silhouettes, the neutron flux cut-off was successfully recognized. The neutron beam size was determined as -3.0 cm in diameter. This type of neutron beam silhouette study is not readily possible by other techniques. (author)

  4. Simple correlations for bubble columns and draft tube airlift reactors with dilute alcohol solutions

    OpenAIRE

    Kojić Predrag S.; Tokić Milenko S.; Čolović Radmilo R.; Šijački Ivana M.

    2009-01-01

    Simple empirical correlations were developed to predict gas holdup, liquid circulation time, downcomer liquid velocity and volumetric mass transfer coefficient in dilute alcohol solutions in bubble columns and draft tube airlift reactors with single orifice sparger. Also, new experiments were conducted with diluted alcohol solutions to n-octanol, expanding the experimental data from C1 up to C8. The proposed empirical correlations include, beside the superficial gas velocity, the alcohol chai...

  5. Experiments and modelling of a draft tube airlift reactor operated at high gas throughputs

    OpenAIRE

    Colombet, Damien; Cockx, Arnaud; Guiraud, Pascal; Legendre, Dominique

    2013-01-01

    One-dimensional modelling of global hydrodynamics and mass transfer is developed for an annulus sparged draft tube airlift reactor operating at high gas throughputs. In a first part, a specific closure law for the mean slip velocity of bubbles in the riser is proposed according for, in one hand, the collective effects on bubble rise velocity and, in the other hand, the size of the liquid recirculation in the airlift riser. This global hydrodynamics model is found towel explain the global gas ...

  6. High-speed roll-to-roll manufacturing of graphene using a concentric tube CVD reactor

    Science.gov (United States)

    Polsen, Erik S.; McNerny, Daniel Q.; Viswanath, B.; Pattinson, Sebastian W.; John Hart, A.

    2015-05-01

    We present the design of a concentric tube (CT) reactor for roll-to-roll chemical vapor deposition (CVD) on flexible substrates, and its application to continuous production of graphene on copper foil. In the CTCVD reactor, the thin foil substrate is helically wrapped around the inner tube, and translates through the gap between the concentric tubes. We use a bench-scale prototype machine to synthesize graphene on copper substrates at translation speeds varying from 25 mm/min to 500 mm/min, and investigate the influence of process parameters on the uniformity and coverage of graphene on a continuously moving foil. At lower speeds, high-quality monolayer graphene is formed; at higher speeds, rapid nucleation of small graphene domains is observed, yet coalescence is prevented by the limited residence time in the CTCVD system. We show that a smooth isothermal transition between the reducing and carbon-containing atmospheres, enabled by injection of the carbon feedstock via radial holes in the inner tube, is essential to high-quality roll-to-roll graphene CVD. We discuss how the foil quality and microstructure limit the uniformity of graphene over macroscopic dimensions. We conclude by discussing means of scaling and reconfiguring the CTCVD design based on general requirements for 2-D materials manufacturing.

  7. Shot navigation for North Carolina barrier island ground penetrating radar collected by East Carolina University in 2002 (ilgpr2002_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  8. Shot navigation for North Carolina barrier island ground penetrating radar collected by East Carolina University in 2005 (ilgpr2005_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  9. Ground Penetrating Radar (GPR) trackline navigation collected by East Carolina University along the North Carolina barrier islands in 2001 (ilgpr2001_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  10. JPEG Images of Ground Penetrating Radar (GPR) data collected by East Carolina University along North Carolina Outer Banks 2002-2005

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  11. Ground Penetrating Radar (GPR) trackline navigation collected by East Carolina University along the North Carolina barrier islands in 2005 (ilgpr2005_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  12. Shot navigation for North Carolina barrier island ground penetrating radar collected by East Carolina University in 2001 (ilgpr2001_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  13. Ground Penetrating Radar (GPR) trackline navigation collected by East Carolina University along the North Carolina barrier islands in 2002 (ilgpr2002_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  14. Optimization study for an epithermal neutron beam for boron neutron capture therapy at the University of Virginia Research Reactor

    International Nuclear Information System (INIS)

    The non-surgical brain cancer treatment modality, Boron Neutron Capture Therapy (BNCT), requires the use of an epithermal neutron beam. This purpose of this thesis was to design an epithermal neutron beam at the University of Virginia Research Reactor (UVAR) suitable for BNCT applications. A suitable epithermal neutron beam for BNCT must have minimal fast neutron and gamma radiation contamination, and yet retain an appreciable intensity. The low power of the UVAR core makes reaching a balance between beam quality and intensity a very challenging design endeavor. The MCNP monte carlo neutron transport code was used to develop an equivalent core radiation source, and to perform the subsequent neutron transport calculations necessary for beam model analysis and development. The code accuracy was validated by benchmarking output against experimental criticality measurements. An epithermal beam was designed for the UVAR, with performance characteristics comparable to beams at facilities with cores of higher power. The epithermal neutron intensity of this beam is 2.2 x 108 n/cm2 · s. The fast neutron and gamma radiation KERMA factors are 10 x 10-11cGy·cm2/nepi and 20 x 10-11 cGy·cm2/nepi, respectively, and the current-to-flux ratio is 0.85. This thesis has shown that the UVAR has the capability to provide BNCT treatments, however the performance characteristics of the final beam of this study were limited by the low core power

  15. Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

    International Nuclear Information System (INIS)

    Heating tests using 53 wt % U3O8-Al pellets show that an exothermic reaction occurs between 875 and 10000C and takes 10 to 20 seconds to reach maximum temperature. The maximum temperature is a function of particle size of the U3O8 with large particles exhibiting lower peak temperatures. The calculated energy release was 123 cal/g of U3O8-aluminum fuel. Tests using aluminum clad outer fuel tube sections gave lower peak temperatures than for pellets. No violent reactions occurred. The results are reasonably consistent with recent reported data indicating that the exothermic U3O8-Al reaction is not an important energy source. The compressive and tensile strengths of U3O8 tubes above 6600C are low. In compression, sections with 2 psi average axial stress failed at 9170C, while sections with 7 psi failed at 6690C. Tubes with U-Al alloy cores failed at about 6700C with no applied load. The stresses in fuel tubes during a reactor transient may range up to several hundred psi and are less than 7 psi only in the upper part of the fuel tube

  16. Virginia Tech among eight schools working to increase underrepresented students pursuing science and engineering

    OpenAIRE

    Owczarski, Mark

    2007-01-01

    Virginia Tech is among eight Virginia and North Carolina colleges and universities working together on a new National Science Foundation sponsored program to increase the number of students from underrepresented groups who pursue degrees in science, technology, engineering and mathematics.

  17. Advanced eddy current technique for measurement of annular gap between pressure tube and calandria tube in Indian Pressurized Heavy Water Reactors (PHWRs)

    International Nuclear Information System (INIS)

    In Indian Pressurised Heavy Water Reactors (PHWRs), the PT (pressure tube) is designed to be nominally concentric with the encircling CT (calandria tube). Due to various factors PT becomes eccentric with respect to CT over the life of reactor. If this becomes excessive, hot PT will come in contact with cold CT. Such a cold spot could act as potential location for initiating blister formation and premature failure of PT. Hence it is important to periodically measure annular gap between PT and CT. An advanced eddy current technique has been successfully developed and incorporated in BARCIS (BARC Channel Inspection System) for measurement of PT-CT gap. (author). 4 refs., 3 figs

  18. Estimation of reactivity effect of neutron beam tube in research reactor through two-dimensional transport calculation. Comparison of radial and tangential beam tubes

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, Masatoshi

    1988-12-01

    The reactivity effects of neutron beam tubes in a research reactor were investigated with the two-dimensional transport code DORT. The core model for the calculation was a two-dimensional cylinder. The reactivity effects of one radial and two tangential beam tubes were estimated through the results of the two-dimensional calculation using the space-dependent weight function which is as defined a product of the macroscopic scattering cross section, the forward neutron flux, the adjoint neutron flux and the volume. The reactivity effect of the tangential beam tube is larger than that of the radial tube. An aluminum wall of a beam tube decreases the reactivity of the core due to the neutron absorption.

  19. Uses of Plutonium Fuel in Pressure-Tube-Type, Heavy-Water-Moderated Thermal Reactors

    International Nuclear Information System (INIS)

    In 1962, a feasibility study was begun in the JAERI on the uses of various nuclear fuels for pressure-tube-type, heavy-water-moderated thermal reactors. This study began with analysis of the use of uranium in heavy-water-moderated thermal reactors such as the CANDU-PHW, CANDU-BLW, SGHW, EL-4, and Ref. 15, D and E lattices, which is designed in the JAERI, from the standpoint of the core design. Then, the ways of using plutonium fuel in the same types were investigated using WATCHTOWER, FLARE and VENUS codes, including: (1) direct substitution of the plutonium from light-water reactors or Magnox reactors, (2) recycle use of the plutonium from heavy-water-moderated reactors, (3) plutonium self-sustaining cycle, and (4) plutonium phoenix fuel. The following conclusions are reported: (1) In the direct substitution of plutonium, somewhat depleted plutonium is more suitable for core design than the plutonium from Magnox reactors or light-water reactors, because the increase in the initial reactivity due to large plutonium absorption cross-section must be prevented. (2) In the plutonium self-sustaining cycle, the fuel burn-up of about 15 000 ∼20000 MWd/t would be expected from natural uranium, and the positive void reactivity which always occurs in the uraniumloaded SGHW or CANDU-BLW lattices is greatly reduced, the latter property giving some margin to bum-out heat flux. (3) It may be concluded from the fuel cycle analysis that the plutonium self-sustaining cycle is equivalent to using slightly enriched uranium (about 1.0 at.%). It may be concluded that the use of plutonium in heavy-water-moderated reactors is technologically feasible and economically advantageous. (author)

  20. Do existing research reactors teach us all about beam tube optimization?

    International Nuclear Information System (INIS)

    The contribution makes the attempt to analyse the data base available in the literature and in Siemens' own projects and to find out potential systematics from the existing research reactor with beam tubes, separated into reactors with different reflectors and distinguished for tangential and radial tubes and cold neutron sources, resp. Some generic calculations serve as gauging data. The contribution is not meant as critics on any design.The results might serve supporting designers and operators when evaluating the pros and cons of existing or planned design in terms of the optimum beam tubes. Existing lacks of systematics are evaluated in view of suitable explanations and constraints, which do not allow optimisation. Examples pf such constraints are the different material layers between fuel zone and reflector zone which have various reasons. The limited data in the literature plus the numerous lacks of precision of the representation of those data should be an incentive to improve the performed analysis by collecting more exact data and re-doing the evaluation before answering the title-question really. (author)

  1. Elasto-plastic behaviour of the pressure tube (Zr-2.5Nb%) of CANDU reactor

    International Nuclear Information System (INIS)

    To ensure the structural integrity and meet the safety conditions in the pressure tubes of Cernavoda CANDU type reactor, subjected to a severe thermodynamic operational environment (operation temperature 560 K - 585 K, tangential stress coefficient σe = 110-130 MPa), to the corroding ambient and to the radiation field (a fast neutron flux of 1017 n/m2s), the knowledge of mechanical parameter evolution, along the lifetime span (about 30 - 40 years), is needed. The present work reports the materials data of samples extracted from a pressure tube of Cernavoda reactor (alloy-Zr97.5Nb2.5), axially submitted, to ambient and 300 deg. C temperatures. The thermodynamic stress monitoring and the experimental data acquisition and processing were carried out by an analog-to-digital converter. The experimental data obtained by this procedure were correlated rather well by means of the Hsu-Bertels law, adequate to describe the mechanical fatigue of elasto-plastic materials. The parameters determined are used in computing and predicting the behaviour of CANDU pressure tube in normal operation conditions

  2. Feasibility study for the application of computed tomography to Savannah River reactor-fuel tubes

    Energy Technology Data Exchange (ETDEWEB)

    Roberson, G.P.; Martz, H.E.; Hester, L.O. [Lawrence Livermore National Lab., CA (United States); Griffin, J.C.; McElroy, R.D. Jr. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1991-12-01

    Gamma-ray computed tomography (CT) is a potentially powerful nondestructive method for assessing enriched uranium-aluminum (U-Al) reactor-fuel tubes produced at the Westinghouse Savannah River Company (WSRC). Several proof-of-principle experiments were performed to obtain data that can be used to determine the feasibility and general operating parameters of a CT instrument for this application. We inspected a mock-fuel tube (contains U-238 instead of U-235) supplied by WSRC to assess the ability of CT to measure and distinguish changes in U density in the U-Al core, to distinguish the end of the U-Al core from Al, and to distinguish changes in the Al cladding thickness. The results indicate that CT can provide meaningful information about interior details and spatial measurements for both the U-Al core and Al cladding of the reactor-fuel tube. This feasibility study demonstrates the quantitative possibilities obtained from the nuclear-spectroscopy-based CT scanners used. 21 refs.

  3. Manipulator for inspection and possible repair of the tubes of heat exchangers, especially of steam generators for nuclear reactors

    International Nuclear Information System (INIS)

    Manipulator for inspecting and, if required, repairing the tubes of heat exchangers, especially for nuclear reactors, comprises a tube bundle set in a tube sheet leading into a steam generator chamber and a manipulator is brought in through a lead-in nozzle. An inspection arm can be inserted and removed through the lead-in nozzle and the nozzle can be closed off tight by a blind flange. The inspection arm comprises a guide tube supported in the leadin nozzle, and a swivel arm supported at the end of the guide tube so that it can rotate in a plane parallel to the tube sheet. The swivel arm carries at its outer end an extendable and retractable mouthpiece carrier with a mouthpiece which can be aligned onto the tube openings. The outer contour of the swivel arm and the mouthpiece carrier in the stretched-out transport position is not greater than the inner contour of the lead-in nozzle

  4. 78 FR 16816 - Television Broadcasting Services; Hampton-Norfolk, Virginia; Norfolk, Virginia-Elizabeth City...

    Science.gov (United States)

    2013-03-19

    ... From the Federal Register Online via the Government Publishing Office FEDERAL COMMUNICATIONS COMMISSION 47 CFR Part 73 Television Broadcasting Services; Hampton-Norfolk, Virginia; Norfolk, Virginia..., North Carolina, and to modify its television station, WHRO-TV's license to specify Norfolk,...

  5. Optimization study for an epithermal neutron beam for boron neutron capture therapy at the University of Virginia Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Burns, T.D. Jr.

    1995-05-01

    The non-surgical brain cancer treatment modality, Boron Neutron Capture Therapy (BNCT), requires the use of an epithermal neutron beam. This purpose of this thesis was to design an epithermal neutron beam at the University of Virginia Research Reactor (UVAR) suitable for BNCT applications. A suitable epithermal neutron beam for BNCT must have minimal fast neutron and gamma radiation contamination, and yet retain an appreciable intensity. The low power of the UVAR core makes reaching a balance between beam quality and intensity a very challenging design endeavor. The MCNP monte carlo neutron transport code was used to develop an equivalent core radiation source, and to perform the subsequent neutron transport calculations necessary for beam model analysis and development. The code accuracy was validated by benchmarking output against experimental criticality measurements. An epithermal beam was designed for the UVAR, with performance characteristics comparable to beams at facilities with cores of higher power. The epithermal neutron intensity of this beam is 2.2 {times} 10{sup 8} n/cm{sup 2} {center_dot} s. The fast neutron and gamma radiation KERMA factors are 10 {times} 10{sup {minus}11}cGy{center_dot}cm{sup 2}/n{sub epi} and 20 {times} 10{sup {minus}11} cGy{center_dot}cm{sup 2}/n{sub epi}, respectively, and the current-to-flux ratio is 0.85. This thesis has shown that the UVAR has the capability to provide BNCT treatments, however the performance characteristics of the final beam of this study were limited by the low core power.

  6. New inverted hydride fuel design concept for pressure tube type super critical water reactors

    International Nuclear Information System (INIS)

    In this study, an innovative core design having inverted configuration has been proposed for pressure tube type supercritical water reactors. In this design the relative positions of fuel and coolant have been inverted and U-Th-Zr-hydride fuel has been used. A coupled neutronics and thermal hydraulics analysis was done for the proposed Inverted Pressure Tube Type (IPTT) SCWR. The neutronics analysis was carried out by using a 3D fine mesh diffusion theory code and thermal hydraulics calculations were done by using single channel model. These two codes were coupled with each other by a link code. The average outlet temperature for the proposed IPTT-SCWR was found to be 625degC with maximum clad surface temperature (MCST) under the design limits i.e. below 850degC. Moreover a core loading pattern has also been proposed to achieve uniform radial power distribution and lower cladding surface temperature. (author)

  7. A computational study on instrumentation guide tube failure during a severe accident in boiling water reactors

    International Nuclear Information System (INIS)

    This paper focuses on the nature and timing of Instrumentation Guide Tube (IGT) failure in case of severe core melt accident in a Nordic type Boiling Water Reactor (BWR). First, a 2D structural analysis of a RPV lower head is performed to determine global vessel deformation, timing and mode of failure. Next, a structural analysis is also performed on a 3D IGT section taking into account the influence of global vessel deformation and thermo-mechanical load from the melt pool. We have found that the IG tube was not clamped in the housing at the time when welding ring of the IGT nozzle has been melted and global failure of the vessel wall has not started yet. This suggests that IGT failure is the dominant failure mode in the considered case of a large (~200 tons) melt pool. (author)

  8. Analysing The Thermalhydraulic Parameters Of VVER-1000 Reactor For The Accident Of Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    To ensure the safety operation of nuclear power plant (NPP), a lot of postulated accident scenarios were considered and analysed. This research chose and analysed the accident of steam generator tube rupture (SGTR) under the actual plant conditions by using the simulation program PCTRAN. The SGTR accident is happen when the NPP is under operation with the steady state condition (power of 3000 MWth, primary pressure of 157 bar and secondary pressure of 63 bar). The accident is initiated by creating a break with equivalent diameter of 100 mm in the area of lower row heat exchanging tubes. The result of analysis is compared with the calculation of the Shariz University, Iran using the thermal hydraulics code RELAP5/mod3.2 and the report in the PSAR data of VVER-1000. This comparison shows that it is possible for using PCTRAN to analyse accidents of VVER-1000 reactor. (author)

  9. Preliminary evaluation of steam generator tube rupture (SGTR) accident in lead cooled reactor

    International Nuclear Information System (INIS)

    In this paper some contributions are provided to the development of a European Lead-cooled System, known as the ELSY project (within EU-6 Framework Project); that will constitute a possible reference system for a large lead-cooled reactor of GEN IV. Steam generator (SG) tubing of this system type might be subject to a variety of degradation processes, such as cracking, wall thinning and potential leakage or rupture, eventually leading to the failure of one or more SG tubes that constitute a steam generator tube rupture (SGTR) accident with possible consequences for the safety of the primary systems. It is therefore of interest for the designer to know how the SG itself, as well as the vessel and internals structures, behave under impulsive loading conditions (in form of a rapid and strong increase of pressure) that can arise as consequences of the interaction between the primary and secondary coolants (lead-water interaction). The analysed initiator event, as already mentioned, is a large break (up to a double ended guillotine break) of one (or more) SG cooling tubes that may become severe enough to determine dangerous effects on the interested structures. In order to better simulate and perform the mentioned postulated SGTR accident sequence analyses, an appropriate numerical model with the available computing resources (FEM codes) was set up at the DIMNP of Pisa University. That model was used to evaluate the effects of the propagation of the blast pressure waves inside the SG structures, taking into account also the sloshing phenomenon that could be induced by the lead primary coolant motions. Therefore the SGTR effects study may be considered as a transient and non linear problem the solution of which provides the 'time histories' of hydrodynamic pressures and stresses on the reactor pressure vessel and internals walls. (author)

  10. Simple correlations for bubble columns and draft tube airlift reactors with dilute alcohol solutions

    Directory of Open Access Journals (Sweden)

    Kojić Predrag S.

    2009-01-01

    Full Text Available Simple empirical correlations were developed to predict gas holdup, liquid circulation time, downcomer liquid velocity and volumetric mass transfer coefficient in dilute alcohol solutions in bubble columns and draft tube airlift reactors with single orifice sparger. Also, new experiments were conducted with diluted alcohol solutions to n-octanol, expanding the experimental data from C1 up to C8. The proposed empirical correlations include, beside the superficial gas velocity, the alcohol chain length as the only factor to characterize the liquid phase. The suggested correlations have shown good agreement between the calculated and the experimental data.

  11. Simulating transients in the subcritical reactor by using a sealed-tube neutron generator

    International Nuclear Information System (INIS)

    Research on transients is one of the main focuses in the field of accelerator driven subcritical reactor (ADSR). But inducing the transients by accelerator is inconvenient and costly. Simulating transients by sealed tube neutron generator, through improving the driver of Penning ion source and modulating trigger signals of pulsed neutron, is a fast, accurate, reliable and low-cost method. This method has been used in simulating transients on the physics test platform of ADSR in China Institute of Atomic Energy (CIAE). And this method also can be applied in PGNAA. (author)

  12. Factors in the selection of broiler tube materials for a civil fast reactor

    International Nuclear Information System (INIS)

    This paper briefly considers some of the factors which must be balanced in the selection of a boiler tube material for a Civil Fast Reactor. The merits and possible demerits of low alloy ferritic steels and the austenitic Alloy 800 are compared with respect to waterside corrosion resistance, mechanical properties, fabrication and weldability and possible effects of exposure to the sodium environment under normal and fault conditions. It is pointed out that although there is operational experience of most of the materials in boiler superheater applications there is little or none in evaporative regimes. (author)

  13. Advanced nondestructive examination of the reactor vessel head penetration tube welds

    International Nuclear Information System (INIS)

    Beside a referent code examination requirements, appearance of the service induced flaws on the Reactor Vessel Head (RVH) penetration tube welds forced development of remotely operated examination tools and techniques. Several systems were developed for examination of RVH PWR type while only one system for examination of VVER - 440 type RVH has been developed by Inetec. In this article the most advanced RVH VVER - 440 type examination techniques such as ultrasonic, eddy current and visual testing techniques as well as remotely operated tool are described. (author)

  14. Remote field eddy current technique for gap measurement of horizontal flux detector guide tube in pressurized heavy water reactor

    International Nuclear Information System (INIS)

    The fuel channels including the pressure tube(PT) and the calandria tube(CT) are important components of the pressurized heavy water reactor(PHWR). A sagging of fuel channel increases by heat and radiation exposure with the increasing operation time. The contact of fuel channel to the Horizontal flux Detector(HFD) guide tube is needed for the power plant safety. In order to solve this safety issue, the electromagnetic technique was applied to measure the status of the guide tube. The Horizontal flux Detector(HFD) guide tube and the Calandria tube(CT) in the Pressurized Heavy Water Reactor(PHWR) are cross-aligned horizontally. The remote field eddy current(RFEC) technology is applied for gap measurement between the HFD guide tube and the CT HFD guide tube can be detected by inserting the RFEC probe into pressure tube(PT) at the crossing point directly. The RFEC signals using the volume integral method(VIM) were simulated for obtaining the optimal inspection parameters. This paper shows that the simulated eddy current signals and the experimental results in variance with the CT/HFD gap.

  15. Oxidation and deuterium uptake of Zr-2.5Nb pressure tubes in CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian deuterium uranium pressurized heavy-water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm2 hydrogen per year) . The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes. (author)

  16. Investigation of hydrodynamic parameters in a draft tube reactor using radioisotope based techniques and conventional method

    International Nuclear Information System (INIS)

    In this study, various techniques were attempted to investigate flow dynamics in the enclosed reactor and the results from the techniques were compared. Radioactive particle tracking (RPT) and industrial single photon emission computed tomography (SPECT) were carried out and the circulation times from them showed a deviation of 17.5%. The circulation time of the RPT was longer than that of SPECT, and it is speculated that the physical dimension of the/ fabricated radioactive particle creates the discrepancy. Particle image velocimetry (PIV) measurements and computational fluid dynamic (CFD) simulations were conducted. The velocity patterns from them were similar to each other in the entire reactor region except near the propeller installed at the bottom of the reactor. - Highlights: • Radioisotope based techniques were used to investigate flow dynamics in a draft tube reactor. • The circulation time measured using RPT is longer than the circulation time measured by the SPECT. • Radioactive particle tracking technique is expected to be helpful to investigate flow dynamics

  17. Safety Assessment of Pressure-Tube Heavy-Water Reactors by Probability Methods

    International Nuclear Information System (INIS)

    The use of probability methods for the assessment of safety has been developed in the United Kingdom for gas-cooled reactors; the development of these methods has proceeded in parallel with a move towards the formulation of quantitative safety criteria; some possible criteria are described. In particular, a method has been developed which is of value for a rapid assessment of a preliminary design in order to reveal potential points of weakness before the design is finalized. The method is of general application and in this paper it is applied to the design of hypothetical 250-MW(e) reactors of both the indirect and direct cycle types in order to illustrate the use of this method for pressure-tube heavy-water reactors and also to provide a comparison of the safety of this reactor system and others, such as the United Kingdom advanced gas-cooled type. A difficulty in the use of probability methods at present is the scarcity of data on failure rates for structures and large items of plant. An essential feature of the method described is a perturbation of the assumed failure rates by factors of order 100, to assess the effect of such uncertainties. The effect of major changes in other features important to safety, such as reliability of the containment systems, is also examined. (author)

  18. The Economical Application of Non-Destructive Testing to Reactor Components, Especially Jacket Tubing

    International Nuclear Information System (INIS)

    The ideal reactor design, in addition to its other desirable characteristics, would require no non-destructive testing. This ideal, like others, will probably never be attained. In any reactor design where cost is an important factor, the question of whether components can be economically tested should be proposed at the same time that questions of fabricability are being considered. Some development of these points as well as a discussion of the importance of non- destructive tests in specification writing is included in this section. Responsibility also rests on the fabricator to use the help provided by non-destructive testing in maintaining quality in the product through various stages in the fabrication process, and to use the test results to indicate those steps in the process most likely to introduce defects in the component. Often it develops that non-destructive testing in earlier stages of component fabrication cannot be replaced economically, if at all, by inspection of the component in finished or semi-finished form. Examples are cited to illustrate this point, particularly with regard to tubing for fuel jacket and heat-exchanger applications. The application of various non-destructive tests during a tube-fabrication development programme is described in some detail. The fabrication and inspection costs for some tubing used for jacket applications by Argonne National Laboratory are compared. Although component inspection in finished form can be minimized by these procedures, it cannot in all cases be eliminated entirely. The economical testing of plates and tubes, especially the latter, is discussed in detail. The discussion is centred around components of stainless steel, Zircaloy, and certain refractory metal alloys. It is shown through various examples that although the use of radiography and penetrants may be useful or even essential steps in the testing, critical inspection of thin-wall tubing must usually be made by either an ultrasonic or an

  19. Study of the U3O8-Al thermite reaction and strength of reactor fuel tubes

    International Nuclear Information System (INIS)

    Research and test reactors are presently operated with aluminum-clad fuel elements containing highly enriched uranium-aluminum alloy cores. To lower the enrichment and still maintain reactivity, the uranium content of the fuel element will need to be higher than currently achievable with alloy fuels. This will necessitate conversion to other forms such as U3O8-aluminum cermets. Above the aluminum melting point, U3O8 and aluminum undergo an exothermic thermite reaction and cermet fuel cores tend to keep their original shape. Both factors could affect the course and consequences of a reactor accident, and prompted an investigation of the behavior of cermet fuels at elevated temperatures. Tests were carried out using pellets and extruded tube-sections with 53 wt % U3O8 in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 9000C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U3O8-aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 6600C. In tension, sections failed at about the aluminum melting point. In compression with 2-psi average axial stress, failure occurred at 9170C, while 7 psi average axial stress produced failure at 6690C

  20. Analysis of transient dry patch behavior on CANDU reactor calandria tubes in a LOCA with late stagnation and impaired ECI

    International Nuclear Information System (INIS)

    An analytical method to describe the behavior of transient dry patches on CANDU reactor calandria tubes has been developed. Dry patches may form following the sagging of a pressure tube onto a calandria tube in certain low-probability scenarios in which a loss-of-coolant accident occurs with subsequent failure or impairment of the emergency cooling injection function. Results of the analysis show that the dry patches will not grow beyond a few degrees on each side of the bottom of the calandria tube and will rewet within a few tens of seconds, with the values depending on the specific CANDU reactor design and the mechanism of dry patch formation and rewetting. Maximum local calandria tube temperatures reached during the transient will be about 5500C to 7000C. There will be no significant effects (0C) on fuel, sheath and maximum pressure tube temperatures. The analytical results provide confidence that pressure tube and calandria tube integrity will not be threatened by dry-patch formation in the LOCA scenarios studied

  1. Present and future beam tube experiments at the 250 kW TRIGA Mark II reactor Wien

    International Nuclear Information System (INIS)

    The four beam tubes and the thermal column at the TRIGA reactor Wien were well used in the reporting period. Since the thermal column is used as a gamma source for different irradiation experiments and as a neutron source for radiography, the other facilities are mainly used for neutron spectroscopy experiments: polarized neutrons, neutron interferometry, small angle scattering and neutron choppers, In the piercing beam tube a fast rabbit system is installed which is mainly used for high precision activation analysis. (author)

  2. JPEG images of chirp seismic data from a 2005 nearshore survey collected by Virginia Institute of Marine Science

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  3. JPEG images of chirp seismic data from a 2002 nearshore survey collected by Virginia Institute of Marine Science

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  4. Pressurized water reactor vessel internals guide tube guide card wear aging management

    International Nuclear Information System (INIS)

    In order for the pressurized water reactors to qualify for life extension, they have to meet the requirements of MRP-277 (Reference 1). MRP-227 lists the various reactor internals components that need to be inspected in order for a plant to qualify for life extension; the upper internals guide tube guide cards are one such component. Aggressive guide card wear in a plant can lead to violation of plant technical specifications, safety issues in the event of insertion, failure of one or more rod cluster control assemblies, or even result in plant shutdown or outage extensions. Owing to the criticality of the guide card wear, as discussed above, the Pressurized Water Reactor Owners Group (PWROG) initiated a guide card wear measurement project, led by Westinghouse. Under this program, wear of the guide cards at three identified lead plants was observed and measured, and, an engineering review of the wear data was completed, and plant specific recommendations based on the engineering reviews was provided. To support this program, Westinghouse also developed criteria to prevent or mitigate guide card wear, which governs the guide card wear measurement. In addition, preliminary root cause analysis was performed for one of the aggressively wearing plants, where some wear aggravating causes and mitigating techniques were determined. Therefore, this paper will discuss Westinghouse's guide card wear criteria and measurement technique, guide card wear trends obtained from measurements conducted in the guide card wear program, possible guide card wear aggravating causes and guide card wear mitigating techniques. (author)

  5. Method of experimental and theoretical modeling for multiple pressure tube rupture for RBMK reactor

    International Nuclear Information System (INIS)

    The rupture of single RBMK reactor channels has occurred at a number of stations with a variety of initiating events. It is assumed in RBMK Safety Cases that the force of the escaping fluid will not cause neighbouring channels to break. This assumption has not been justified. A chain reaction of tube breaks could over-pressurise the reactor cavity leading to catastrophic failure of the containment. To validate the claims of the RBMK Safety Cases the Electrogorsk Research and Engineering Centre, in participation with experts from the Institute of Mechanics of RAS, has developed the method of interacting multiscale physical and mathematical modelling for coupled thermophysical, hydrogasodynamic processes and deformation and break processes causing and (or) accompanying potential failures, design and beyond the design RBMK reactor accidents. To realise the method the set of rigs, physical and mathematical models and specialized computer codes are under creation. This article sets out an experimental philosophy and programme for achieving this objective to solve the problem of credibility or non-credibility for multiple fuel channel rupture in RBMK.(author)

  6. Modeling and simulation of tube-shell reactor for dimethyl-ether synthesis from coal-based synthesis gas

    Institute of Scientific and Technical Information of China (English)

    CHEN Da-sheng; ZHANG Hai-tao; YING Wei-yong; FANG Ding-ye

    2011-01-01

    Mathematical simulation was performed on tube-shell reactor for dimethyl ether (DME) synthesis from coal-based syngas. The model was established based on kinetics of dimethyl-ether synthesis from syngas over a bifunctional catalyst,which is mixed by methanol synthesis catalyst and dehydration catalyst as 1:1 mass ratio. Methanol synthesis from CO and CO2 and methanol dehydration were selected as three-independent reactions, CO, CO2, and DME as key components to establish the one-dimensional mathematical model of the reactor. The gas concentration and temperature profiles inside the reactor tubes were obtained. The operating conditions affecting DME synthesis were also discussed based on the model. The simulations indicate that higher pressure and lower temperature at the inlet and rich hydrogen in the reactant are favorable in direct DME synthesis in fixed-bed process, and the temperature of boiling water affect the reactor performance seriously.

  7. Update of operating experience with cold-worked Zr-2.5%Nb pressure tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Zr-2.5 Nb pressure tubes are now used in all CANDU reactors. To ensure they perform reliably, their performance is carefully monitored. Both in situ inspection and sampling and testing techniques for tubes periodically removed from reactors have been developed. The data from these inspections and tests, together with models developed from research programs give confidence that pressure tubes will function effectively and safely for their design life. This presentation will describe how service life affects changes in the major material parameters in pressure tubes and the resulting maintenance activities resulting from those changes. Thermal creep, irradiation creep and irradiation growth change the dimensions during service, and axial elongation due to growth and sag due to creep in the older reactors have resulted in major maintenance programs. However, the dimensional changes continue to follow the behaviour predicted by the design equations and in the newer reactors should not limit service life. Extensive in situ sampling and the analysis of the tubes recently removed from Pickering Unit 3 indicate that hydrogen ingress into the pressure tubes from corrosion on the inside surface is very low and tests on irradiated material indicate that it should continue to remain low. The ingress rate from the annulus gas side can be significant if the integrity of the oxide on the outside surface is not maintained as a barrier. To maintain te integrity of the autoclave oxide, the recommended annulus gas is carbon dioxide, with oxygen addition, and adequate flow must be ensured. An explanation of the cause of relatively high hydrogen concentrations in a few Pickering A Zr-2.5% Nb pressure tubes has been developed defining the role of annulus side ingress. The model developed to predict the time and conditions to initiate blisters in pressure tubes that are in contact with their calandria tubes has been validated by the inspection, removal and examination of tubes and gives

  8. Internal welding of tube-to-tubesheet joints of steam generator for sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    In the steam generator for a sodium-cooled fast breeder reactor, there are many joints of tubes and tube sheets. For the internal welding of small diameter, thick walled tubes and tubesheets, welding method has been developed, which gives high quality welding with good reproducibility. In this method, the pressure of shield gas is controlled suitably, and consideration is given to the composition of the shield gas. As a means to ensure the quality of welds, the technique of internal radiographic test has also been established. Both the welding method and the test were able to be applied successfully to the steam generator of practical size. (Mori, K.)

  9. Combustion, cofiring and emissions characteristics of torrefied biomass in a drop tube reactor

    International Nuclear Information System (INIS)

    The study investigates cofiring characteristics of torrefied biomass fuels at 50% thermal shares with coals and 100% combustion cases. Experiments were carried out in a 20 kW, electrically heated, drop-tube reactor. Fuels used include a range of torrefied biomass fuels, non-thermally treated white wood pellets, a high volatile bituminous coal and a lignite coal. The reactor was maintained at 1200 °C while the overall stoichiometric ratio was kept constant at 1.15 for all combustion cases. Measurements were performed to evaluate combustion reactivity, emissions and burn-out. Torrefied biomass fuels in comparison to non-thermally treated wood contain a lower amount of volatiles. For the tests performed at a similar particle size distribution, the reduced volatile content did not impact combustion reactivity significantly. Delay in combustion was only observed for test fuel with a lower amount of fine particles. The particle size distribution of the pulverised grinds therefore impacts combustion reactivity more. Sulphur and nitrogen contents of woody biomass fuels are low. Blending woody biomass with coal lowers the emissions of SO2 mainly as a result of dilution. NOX emissions have a more complex dependency on the nitrogen content. Factors such as volatile content of the fuels, fuel type, furnace and burner configurations also impact the final NOX emissions. In comparison to unstaged combustion, the nitrogen conversion to NOX declined from 34% to 9% for air-staged co-combustion of torrefied biomass and hard coal. For the air-staged mono-combustion cases, nitrogen conversion to NOX declined from between 42% and 48% to about 10%–14%. - Highlights: • Impact of torrefaction on cofiring was studied at high heating rates in a drop tube. • Cofiring of torrefied biomasses at high thermal shares (50% and higher) is feasible. • Particle size impacts biomass combustion reactivity more than torrefaction. • In a drop tube reactor, torrefaction has no negative impact

  10. Break flow modeling for a steam generator tube rupture (SGTR) incident in a pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    The design-basis steam generator tube rupture (SGTR) scenario for the pressurized water reactor (PWR) postulates an instantaneous double-ended break of a steam generator (SG) U-tube. The flow rate through the broken U-tube depends on the primary-to-secondary side differential pressure in the affected SG, the primary coolant subcooling, and the break location along the U-tube. In this report, the RELAP5/MOD2 code's capability in predicting the SGTR break flow rate is assessed against experiments conducted on the Large Scale Test Facility (LSTF). The code is then used to predict break flow rate in the PWR for typical SGTR situations. It is shown that the code simulates well the break flow rates in the LSTF experiments for both single-phase and two-phase discharges, including two-phase critical flow discharge. The calculated PWR break flow rate takes a maximum for a break occurring at the lower end of the U-tube, on its cold leg side, because of the combined influence of tube-inlet fluid subcooling and frictional pressure drop along the broken tube. Modeling the tube frictional pressure drop is important to predict the break flow rate dependence on inlet fluid sub-cooling; simplified break flow modeling which applies a constant discharge coefficient less than unity, instead of modeling explicitly the tube frictional length, fails to predict the change in break flow rate accurately if the inlet subcooling varies for a wide range. (author)

  11. The Influence of Slight Protuberances in a Micro-Tube Reactor on Methane/Moist Air Catalytic Combustion

    OpenAIRE

    Ruirui Wang; Jingyu Ran; Xuesen Du; Juntian Niu; Wenjie Qi

    2016-01-01

    The combustion characteristics of methane/moist air in micro-tube reactors with different numbers and shapes of inner wall protuberances are investigated in this paper. The micro-reactor with one rectangular protuberance (six different sizes) was studied firstly, and it is shown that reactions near the protuberance are mainly controlled by diffusion, which has little effect on the outlet temperature and methane conversion rate. The formation of cavities and recirculation zones in the vicinity...

  12. Location of vibracores from offshore of Dare County, North Carolina (ncd_cores.shp, geographic, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  13. Chirp trackline navigation from USGS cruise 2004-005-FA from Pamlico Sound, North Carolina (bbc2004005_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  14. Boomer seismic tracklines from USGS cruise 2003-005-FA from Pamlico Sound, North Carolina (bbb2003005_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  15. Chirp shotpoint navigation from USGS cruise 2002-015-FA from Pamlico Sound, North Carolina (bbc2002015_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  16. Boomer shotpoint navigation from USGS cruise 2002-015-FA from Pamlico Sound, North Carolina (bbb2002015_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  17. Chirp shotpoint navigation from USGS cruise 2001-013-FA from Pamlico Sound, North Carolina (bbc2001013_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  18. Boomer seismic tracklines from USGS cruise 2004-005-FA from Pamlico Sound, North Carolina (bbb2004005_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  19. Boomer seismic tracklines from USGS cruise 2003-042-FA from Pamlico Sound, North Carolina (bbb2003042_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  20. Chirp shotpoint navigation from USGS cruise 2003-042-FA from Pamlico Sound, North Carolina (bbc2003042_shot.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  1. Boomer shotpoint navigation from USGS cruise 2003-005-FA from Pamlico Sound, North Carolina (bbb2003005_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  2. Boomer seismic tracklines from USGS cruise 2004-006-FA from Pamlico Sound, North Carolina (bbb2004006_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  3. Chirp navigation tracklines from USGS cruise 2002-015-FA from Pamlico Sound, North Carolina (bbc2002015_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  4. Chirp navigation tracklines from USGS cruise 2001-013-FA from Albemarle Sound, North Carolina (bbc2001013_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  5. Chirp shotpoint navigation (from USGS cruise 2003-005-FA from Pamlico Sound, North Carolina (bbc2003005_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  6. Boomer shotpoint navigation from USGS cruise 2004-005-FA from Pamlico Sound, North Carolina (bbb2004005_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  7. Location of vibracores collected from nearshore off of Duck, North Carolina in 2005 (vims_cores.shp, geographic, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  8. Boomer shotpoint navigation from USGS cruise 2003-042-FA from Pamlico Sound, North Carolina (bbb2003042_shot200.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  9. Boomer shotpoint navigation from USGS cruise 2004-006-FA from Pamlico Sound, North Carolina (bbb2004006_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  10. bbc2004005_shots.shp: Chirp shotpoint navigation from USGS cruise 2004-005-FA from Pamlico Sound, North Carolina

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  11. Boomer seismic navigation from USGS cruise 2002-015-FA from Pamlico Sound, North Carolina (bbb2002015_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  12. Boomer shotpoint navigation from USGS cruise 2001-013-FA from Albemarle Sound, North Carolina (bbb2001013_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  13. Sizing cracks in thin-walled CANDU reactor pressure tubes using crack-tip diffraction

    International Nuclear Information System (INIS)

    The most practical nondestructive means of measuring the depth of cracks approximately 0.4 mm deep in CANDU reactor pressure tubes is the ultrasonic crack-tip diffraction method. Initially, optimum ultrasonic parameters for wave mode, transducer frequency, main-bang pulse characteristics, incident and diffracted angles were obtained on three fatigue cracks, based on the criteria of maximum signal amplitude and accuracy in determination of crack depth. In addition, three signal processing techniques, auto and cross-correlation, rectification and smoothing and the magnitude of the analytic signal, were used to obtain time measurements. The results of these measurements are presented. Except for the first fatigue crack, the depth calculations were accurate to within the specified range of ± 0.1 mm

  14. Control of alkaline stress corrosion cracking in pressurized-water reactor steam generator tubing

    International Nuclear Information System (INIS)

    Outer-diameter stress corrosion cracking (ODSCC) of alloy 600 (UNS N06600) tubings in steam generators of the Kori-1 pressurized-water reactor (PWR) caused an unscheduled outage in 1994. Failure analysis and remedy development studies were undertaken to avoid a recurrence. Destructive examination of a removed tube indicated axial intergranular cracks developed at the top of sludge caused by a boiling crevice geometry. A high ODSCC propagation rate was attributed to high local pH and increased corrosion potential resulting from oxidized copper presumably formed during the maintenance outage and plant heatup. Remedial measures included: (1) crevice neutralization by crevice flushing with boric acid (H3BO3) and molar ratio control using ammonium chloride (NH4Cl), (2) corrosion potential reduction by hydrazine (H2NNH2) soaking and suppression of oxygen below 20 ppb to avoid copper oxide formation, (3) titanium dioxide (TiO2) inhibitor soaking, and (4) temperature reduction of 5 C. Since application of the remedy program, no significant ODSCC has been observed, which clearly demonstrates the benefit of departing from an oxidizing alkaline environment. In addition, the TiO2 inhibitor appeared to have a positive effect, warranting further examination

  15. Postirradiation properties of the 6061-T6 aluminum High Flux Isotope Reactor hydraulic tube

    International Nuclear Information System (INIS)

    A tube of 6061 aluminum alloy in a T6 temper, precipitation-hardened with Mg2Si, was examined after irradiation in the core of the High Flux Isotope Reactor to fluences up to 1.3 x 1023 neutrons (n)/cm2 (0.1 MeV) and 3.1 x 1023 n/cm2 (thermal) in contact with the cooling water at a temperature of about 550C. The alloy displayed up to 2.5 percent swelling due mainly to a precipitate of transmutation-produced silicon of which more than 6 weight percent was formed. Some cavities were also observed. Tension tests in the temperature range 55 to 2000C showed radiation-induced increases in yield stresses and ultimate stresses of 50 to 80 percent; elongation was reduced from the range 10 to 15 percent to about 5 percent at 550C and to about 3 percent at test temperatures above 1000C. The fracture mode was changed from transgranular tearing around inclusions to a mixture of transgranular tearing and ductile intergranular separation. These changes are attributed primarily to the radiation-induced silicon precipitate. A rim of intergranular cracks formed at the originally oxidized surfaces of the tube duringtension testing and became deeper with increasing neutron irradiation and increasing temperature

  16. Postirradiation properties of the 6061-T6 aluminum high flux isotope reactor hydraulic tube

    International Nuclear Information System (INIS)

    A tube of 6061 aluminum alloy in a T6 temper, precipitation-hardened with Mg2Si, was examined after irradiation in the core of the High Flux Isotope Reactor to fluences up to 1.3 x 1023 neutrons (n)/cm2 (0.1 MeV) and 3.1 x 1023 n/cm2 (thermal) in contact with the cooling water at a temperature of about 550C. The alloy displayed up to 2.5 percent swelling due mainly to a precipitate of transmutaion-produced silicon of which more than 6 weight perent was formed. Some cavities were also observed. Tension tests in the temperature range 55 to 2000C showed radiation-induced increases in yield stresses and ultimate stresses of 50 to 80 percent; elongation was reduced from the range 10 to 15 percent to about 5 percent at 55C0 and to about 3 percent at test temperatures above 100C0. The fracture mode was changed from transgranular tearing around inclusions to a mixture of transgranular tearing and ductile intergradular separation. These changes are attributed primarily to the radiation-induced silicon precipitate. A rim of intergranular cracks formed at the originally oxidized surfaces of the tube during tension testing and became deeper with increasing neutron irradiation and increasing temperature

  17. Steam generator and preheater tube ID fouling and the impact on reactor inlet header temperature and eddy current inspections

    International Nuclear Information System (INIS)

    Materials selection is an important consideration in new build and refurbishment of Heavy Water Reactors (HWR). This paper will focus on the impacts of the deposit of magnetite on the tube ID of steam generators and preheater. Bruce Power OPEX is being shared to illustrate the importance of materials selection. The deposit of magnetite on the tube ID of steam generators (SG) and preheater (PH) has two significant impacts that will be presented. Firstly, the degradation in SG and PH thermal performance causes a rise in the reactor inlet header temperature (RIHT). This rising trend continues unabated as long as deposits on the tube ID continues. If not managed this may result in loss of production due to the RIHT limits being reached. Mitigating actions such as tube ID cleaning is only a temporary solution as it does not stop the root cause which is feeder flow accelerated corrosion (FAC). Secondly, deposit of magnetite on the tube ID of steam generators (SG) and preheater (PH) has an impact on tube inspections as required by CSA N285.4. There are two impacts on SG and PH inspections. ID deposits reduces the clearance for eddy current probes in the tubes and make it more difficult to acquire inspection data. Additionally, tube ID deposits can reduce the effectiveness to detect and size flaws in the SG and PH tubes. Both issues make eddy current inspection a challenge for the utilities. These impacts affect the operation and inspection and maintenance of CANDU nuclear power plants at Bruce Power. Where possible these issues should be addressed in any future new build or refurbishment of HWR power plants. (author)

  18. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, K.C. [Univ. of Turabo, Gurabo, Puerto (Puerto Rico). College of Engineering; Yahr, G.T. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K{sub Q} due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail.

  19. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    International Nuclear Information System (INIS)

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in KQ due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail

  20. Design and manufacture of tube to tubesheet joints of steam generator for 500 MWe Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Prototype Fast Breeder Reactor (PFBR) is 500 MWe pool type sodium cooled fast reactor. Presently this reactor is at advanced stage of construction at Kalpakkam. The main function of the steam generator is to extract the reactor heat through secondary sodium system and convert the feed water into superheated steam in the tubes of steam generators. The steam generator is a vertical shell and tube type heat exchanger with liquid sodium in the shell side and water/steam in the tube side. Operating experience of FBRs have shown that steam generator (SG) holds the key to commercial success of such reactors. Tube leakage is a serious problem and the prevention of sodium water reaction incident in the SG is essential to maintain the plant availability. In case of crack/failure in tube, high pressure water/steam reacts with shell side sodium and results in exothermic reaction with evolution of hydrogen, corrosive reaction products and intense local heat depending on leak size. This high reactive nature of sodium with water/steam requires that sodium to water/steam boundaries of steam generators must possess a high degree of reliability against failure. This is achieved in design and manufacturing by maximising the tube integrity and more importantly by proper selection of tube to tubesheet joint configuration. The principal material of construction of SG is Modified 9Cr-1Mo steel. The tubes are seamless and produced by electric arc melting followed by Electro Slag Refining (ESR) with tight control on inclusion content. Ultrasonic and eddy current testing is done on entire tube length in accordance with ASME SEC III Class I. Long seamless tubes (each 23m) are used in order to reduce the number of tube to tubesheet welds.Each SG has 547 tubes and there are 9 SG in the reactor including one spare module. There is no tube to tube joint as the aim is to minimise the number of welds to increase reliability.Tube to tubesheet joint selected for PFBR steam generator is of internal

  1. ISI of reactor vessel tubes of CIRUS and process water-sea water heat exchangers of DHRUVA

    International Nuclear Information System (INIS)

    Full text: CIRUS is a 40 MW(th) research reactor which has been in operation for the last 37 years. Various components were found to be showing signs of ageing and are in need of repair/replacement. A detailed study was undertaken to check the condition of the reactor components to assess their life expectancy. Under the CRP a special technique using eddy current testing (ECT) was developed to examine the condition of the reactor vessel tubes using the available MIZ-17 eddy current test equipment. A test probe and reference standard were developed for ECT examination. After satisfactory trials approval for the test procedure was obtained from the safety authorities. Mr. Ranade presented the work currently under way. He mentioned that CIRUS has been undergoing refurbishment since the end of 1997. Detailed inspection of all the reactor vessel tubes as per the approved procedure is in progress. Results available for the first 28 tubes inspected indicate that they are in a healthy condition. DHRUVA is a 100 MW(th) research reactor which has been operating since 1985. Regular in-service inspections of the process water/sea water heat exchanger tubes are being carried out using ECT methods. Based on these examinations it was felt necessary to carry out a 3-D flow field analysis employing a validated computer code to get insight with simulated flow conditions to detect any zones which may be susceptible to flow induced vibrations. As per the preliminary report of the 3-D flow field analysis, the heat exchanger tubes coming in these zones are being monitored by ECT on a routine basis. Further ET examination will be carried out when the final results of the 3-D flow field analysis have been obtained. (author)

  2. Intercrystalline Stress Corrosion of Inconel 600 Inspection Tubes in the Aagesta Reactor

    International Nuclear Information System (INIS)

    Intercrystalline stress corrosion cracking has occurred in the Aagesta reactor in three so-called inspection tubes made of Inconel 600. The tubes had been exposed to 217 deg C light water, containing 1-4 ppm LiOH (later KOH) but only small amounts of oxygen, chloride and other impurities. Some of the circumferential cracks developed in or at crevices on the outside surface. At these positions constituents dissolved in the water may have concentrated. The crevices are likely to have contained a gas phase, mainly nitrogen. Local boiling in the crevices may also have occurred. Some few cracks were also found outside the crevice region. Irradiation effects can be neglected. No surface contamination could be detected except for a very minor fluoride content (1 μg/cm2). The failed tubes had been subjected to high stresses, partly remaining from milling, partly induced by welding operations. The possibility that stresses slightly above the 0.2 per cent offset yield strength have occurred at the operating temperature cannot be excluded. The cracked tube material contained a large amount of carbide particles and other precipitates, both at grain boundaries and in the interior of grains. The particles appeared as stringers in circumferential zones. Zones depleted in precipitates were found along grain boundaries. The failed tube turned out to have an unusually high mechanical strength, likely due to a combination of some kind of ageing process and cold work (1.0 - 1.3 per cent plastic strain). Laboratory exposures of stressed surplus material in high purity water and in 1 M LiOH at 220 deg C showed some pitting but no cracking after 6800 h and 5900 h respectively. Though the encountered failures may have developed because of influence of some few or several of the above-mentioned detrimental factors, the actual cause cannot be stated with certainty. In the literature information is given concerning intercrystalline stress corrosion cracking of Inconel 600 both in caustic

  3. Development of poison injection code-COPJET for high pressure liquid poison injection in pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Shut Down System-2 (SDS-2) in advanced vertical pressure tube type reactor, provides rapid reactor shutdown by high pressure injection of a neutron absorbing liquid called poison, into the moderator in the calandria. Poison inside the calandria is distributed by poison jets issued from holes provided in the injection tubes. Effectiveness of the system depends on the rate and spread of the poison in the moderator. In this study, a transient one-dimensional (1-D) hydraulic code, COPJET is developed, to predict the performance of system by predicting poison jet length with time. Validation of the COPJET is done with the data available in literature. Thereafter, it is applied for poison jet length prediction of advanced vertical pressure type reactor. (author)

  4. Using 6- and 8-tube IRT-4M fuel assemblies in WWR-SM research reactor core

    International Nuclear Information System (INIS)

    The WWR-SM reactor at the Institute of Nuclear Physics of Academy of Sciences (INP AS) in Uzbekistan was converted to 6-tube IRT-4M LEU (19.7%) fuel. Presently, INP intends to also use IRT-4M 8-tube FA, and a safety analysis for these 'mixed' (8-tube and 6-tube FA) cores is required by the regulatory authorities. This paper presents results of the safety analysis being performed for the 'mixed' cores. Neutronics analysis (burnup, power distributions and shutdown margin), steady-state thermal hydraulics analysis, kinetics parameters for these mixed cores are discussed in this paper. These results will be used to amend the present SAR. (authors)

  5. Control Rod Ejection Accident while Using 6- and 8-Tube IRT-4M Fuel Assemblies in WWR-SM Research Reactor Core

    International Nuclear Information System (INIS)

    The WWR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences (INP AS) in Uzbekistan was converted to 6-tube IRT-4M LEU (19.7%) fuel in 2009. Presently, INP intends to also use IRT-4M 8-tube FA, and a safety analysis for these 'mixed' (8-tube and 6-tube FA) cores is required by the regulatory authorities. This paper presents results of control rod ejection transient analysis for these mixed cores

  6. Theoretical and experimental modeling of the multiple pressure tube rupture for RBMK reactor. Pt. 1

    International Nuclear Information System (INIS)

    Rupture of a single fuel channel (pressure tube) or several fuel channels of the RBMK may occur in service conditions on NPPs with a variety of initiating events. It is assumed in RBMK Safety Cases that the force of the escaping fluid will not cause neighboring channels to break. This assumption has not been justified. Hence, an analysis of the multiple pressure tube rupture (MPTR) possibility is needed. The analysis of the MPTR problem requires performing a series of theoretical and experimental studies of separate physical processes running in the RBMK reactor, as well as development of mathematical models and their physical equivalents. The experimental rigs concerned the MPTR problem have been designed and constructed at Electrogorsk Research and Engineering Center, Russia. Investigation of the circumstances and mechanisms of a single channel rupture at the various conditions and scenarios is one of the main stages of the MPTR problem analysis. Theoretical models of the single channel rupture under thermal and mechanical loading have been developed including a channel constrained by the graphite block. Deformation of the channel under internal pressure and localized thermal action is modeled within the framework of the nonlinear shells theory taking into account physically nonlinear material behavior. Computer program based on these models enables to describe the thermomechanical deformation of a single channel and to predict rupture moment. Theoretical studies were accompanied by experimental modeling single channel rupture by means of series experimental examinations at TKR-F test rig (Model of an Accidental Channel). This test rig represents a model of the single disrupted fuel channel in a surrounding graphite column. Experimental examinations make possible the development and verification of theoretical models and make more exact the conditions and mechanism of a single channel rupture. Theoretical and experimental modeling consolidation sets out technique

  7. Methods of evaluation of accuracy with multiple essential parameters for eddy current measurement of pressure tube to calandria tube gap in CANDU reactors

    International Nuclear Information System (INIS)

    The purpose of inspection qualification of a particular inspection system is to show that it meets applicable inspection specification requirements. Often a requirement of the inspection system is that it meets a particular accuracy. In the case of a system with multiple inputs accompanied by additional influential parameters, calculation of the system's output accuracy can be formidable. Measurement of pressure-tube to calandria tube gap in CANDU reactors using an eddy current based technique is presented as a particular example of a system where multiple essential parameters combine to generate a final uncertainty for the inspection system. This paper outlines two possible methods of calculating such a system's accuracy, and discusses the advantages and disadvantages of each. (author)

  8. Monte Carlo simulation of the thermal column and beam tube of the TRIGA Mark II research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R., E-mail: rustamzia@yahoo.com [Atominstitute (ATI), Vienna University of Technology (TU Wien), Stadion allee 2, A-1020 Vienna (Austria); Karimzadeh, S.; Stummer, T.; Boeck, H. [Atominstitute (ATI), Vienna University of Technology (TU Wien), Stadion allee 2, A-1020 Vienna (Austria)

    2011-08-15

    Highlights: > Neutronics parameters of the reactor shielding. > Biological shielding of the TRIGA reactor. > Thermal flux measurement in the thermal column and BT-A. > MCNP model validation. - Abstract: The Monet Carlo simulation of the TRIGA Mark II research reactor core has been performed employing the radiation transport computer code MCNP5. The model has been confirmed experimentally in the PhD research work at the Atominstitute (ATI) of the Vienna University of Technology. The MCNP model has been extended to complete biological shielding of the reactor including the thermal column, radiographic collimator and four beam tubes. This paper presents the MCNP simulated results in the thermal column and one of the beam tubes (beam tube A) of the reactor. To validate these theoretical results, thermal neutron flux density measurements using the gold foil activation method have been performed in the thermal column and beam tube A (BT-A). In the thermal column, the theoretical and experimental results are in fairly good agreement i.e. maximum thermal flux density in the centre decreases in radial direction. Further, it is also agreed that thermal flux densities in the lower part is greater than the upper part of the thermal column. In the BT-A experiment, the thermal flux density distribution is measured using gold foil. The experimental and theoretical diffusion lengths have been determined as 10.77 cm and 9.36 cm respectively with only 13% difference, reflecting good agreement between the experimental and simulated results. To save the computational cost and to incorporate the accurate and complete information of each individual Monte Carlo MC particle tracks, the surface source writing capability of MCNP has been utilized to the TRIGA shielding model. The variance reduction techniques have been applied to improve the statistics of the problem and to save computational efforts.

  9. Monte Carlo simulation of the thermal column and beam tube of the TRIGA Mark II research reactor

    International Nuclear Information System (INIS)

    Highlights: → Neutronics parameters of the reactor shielding. → Biological shielding of the TRIGA reactor. → Thermal flux measurement in the thermal column and BT-A. → MCNP model validation. - Abstract: The Monet Carlo simulation of the TRIGA Mark II research reactor core has been performed employing the radiation transport computer code MCNP5. The model has been confirmed experimentally in the PhD research work at the Atominstitute (ATI) of the Vienna University of Technology. The MCNP model has been extended to complete biological shielding of the reactor including the thermal column, radiographic collimator and four beam tubes. This paper presents the MCNP simulated results in the thermal column and one of the beam tubes (beam tube A) of the reactor. To validate these theoretical results, thermal neutron flux density measurements using the gold foil activation method have been performed in the thermal column and beam tube A (BT-A). In the thermal column, the theoretical and experimental results are in fairly good agreement i.e. maximum thermal flux density in the centre decreases in radial direction. Further, it is also agreed that thermal flux densities in the lower part is greater than the upper part of the thermal column. In the BT-A experiment, the thermal flux density distribution is measured using gold foil. The experimental and theoretical diffusion lengths have been determined as 10.77 cm and 9.36 cm respectively with only 13% difference, reflecting good agreement between the experimental and simulated results. To save the computational cost and to incorporate the accurate and complete information of each individual Monte Carlo MC particle tracks, the surface source writing capability of MCNP has been utilized to the TRIGA shielding model. The variance reduction techniques have been applied to improve the statistics of the problem and to save computational efforts.

  10. High-definition radiography of tube-to-tubesheet welds of steam generator of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    In the steam generator of the Prototype Fast Breeder Reactor (PFBR), steam is generated by the transfer of heat from secondary sodium to water. Due to the inherent dangers of sodium-water reaction, the integrity of weld joints separating sodium and water/steam is of paramount importance. This is particularly true and very important for the tube-to-tubesheet joints. This paper discusses the use of projective magnification technique by microfocal radiography for the quality evaluation and optimisation of the welding parameters of such small tube-to-tubesheet welds of the steam generator of PFBR. (author)

  11. Development of magnetic flux leakage technique for examination of steam generator tubes of prototype fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: • For non-destructive detection of small localized defects in SG tubes of PFBR, tandem GMR array sensors based MFL technique developed. • 3D-finite element modeling performed for optimization of magnetizing current and spacing between the magnetizing coils. • The optimized magnetizing structure with ferrite core and guides detected 0.54 mm deep OD circumferential notch, 0.56 mm deep flat bottom hole, and 1.08 mm diameter hole in the tube with a SNR better than 6 dB. • Images of notches have been obtained using the tandem GMR array sensor. • The use of MFL and remote field eddy current techniques is expected to ensure comprehensive inspection of SG tubes of PFBR. - Abstract: For non-destructive examination of small diameter (outer diameter, OD 17.2 mm) and thick walled (wall thickness, 2.3 mm) ferromagnetic Modified 9Cr–1Mo steel steam generator (SG) tubes of Prototype Fast Breeder Reactor (PFBR), this paper proposes magnetic flux leakage (MFL) technique. Three dimensional finite element (3D-FE) modeling has been performed to optimize the magnetizing unit and inter-coil spacing of bobbin coils used for axial magnetization of the tube. The performance of the technique has been evaluated experimentally by measuring the axial (Ba) component of the leakage fields from localized machined defects in SG tubes. The MFL technique has shown capability to detect and image tube outside defects with a signal-to-noise ratio (SNR) better than 6 dB. Study reveals that Inconel support plates surrounding the SG tubes do not influence the MFL signals. As the MFL technique can detect localized defects in the presence of support plates as well as sodium and the remote field eddy current technique is sensitive to distributed wall thinning, their combined use will ensure comprehensive inspection of the SG tubes

  12. Chirp navigation tracklines collected by Virginia Institute of Marine Science in 2005 along the nearshore region of the northern Outer Banks, NC (nsc2005_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  13. Chirp shotpoint navigation collected by Virginia Institute of Marine Science along the nearshore region of the northern Outer Banks, NC (nsc2002_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  14. Chirp shotpoint navigation collected by Virginia Institute of Marine Science along the nearshore region of the northern Outer Banks, NC (nsc2005_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  15. Chirp navigation tracklines collected by Virginia Institute of Marine Science in 2002 along the nearshore region of the northern Outer Banks, NC (nsc2002_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  16. Hydrogeological and Groundwater Flow Model for C, K, L, and P Reactor Areas, Savannah River Site, Aiken, South Carolina

    International Nuclear Information System (INIS)

    A regional groundwater flow model encompassing approximately 100 mi2 surrounding the C, K. L. and P reactor areas has been developed. The Reactor flow model is designed to meet the planning objectives outlined in the General Groundwater Strategy for Reactor Area Projects by providing a common framework for analyzing groundwater flow, contaminant migration and remedial alternatives within the Reactor Projects team of the Environmental Restoration Department

  17. Study of hydride blisters grown on Zr-2.5Nb pressure tube spool piece under simulated condition of in-reactor pressure and temperature

    International Nuclear Information System (INIS)

    Indian Pressurised Heavy Water Reactor (PHWR) have pressure tubes, made from zirconium alloy. These pressure tubes undergo corrosion with the high temperature (300 deg C) heavy water coolant under the reactor environment and pick up a part of hydrogen generated as result of this corrosion reaction. This hydrogen affects the integrity of pressure tubes in many ways; nucleation and growth of hydride blisters being one of them. The present study has been carried out to understand the mechanisms of nucleation and growth of hydride blisters and their effect on the serviceability of the component in the reactor environment. (author)

  18. Analyses of the reflector tank, cold source, and beam tube cooling for ANS reactor

    International Nuclear Information System (INIS)

    This report describes my work as an intern with Martin Marietta Energy Systems, Inc., in the summer of 1991. I was assigned to the Reactor Technology Engineering Department, working on the Advanced Neutron Source (ANS). My first project was to select and analyze sealing systems for the top of the diverter/reflector tank. This involved investigating various metal seals and calculating the forces necessary to maintain an adequate seal. The force calculations led to an analysis of several bolt patterns and lockring concepts that could be used to maintain a seal on the vessel. Another project involved some pressure vessel stress calculations and the calculation of the center of gravity for the cold source assembly. I also completed some sketches of possible cooling channel patterns for the inner vessel of the cold source. In addition, I worked on some thermal design analyses for the reflector tank and beam tubes, including heat transfer calculations and assisting in Patran and Pthermal analyses. To supplement the ANS work, I worked on other projects. I completed some stress/deflection analyses on several different beams. These analyses were done with the aid of CAASE, a beam-analysis software package. An additional project involved bending analysis on a carbon removal system. This study was done to find the deflection of a complex-shaped beam when loaded with a full waste can

  19. Results from integral tests of single reformer tubes under simulated nuclear reactor conditions

    International Nuclear Information System (INIS)

    The possibility of supplying high temperature heat from a HTGR for process application is being investigated at some places in the world. In all programmes or projects existing with respect to this application, the endothermic steam reforming of methane is one main step in the transmission of heat produced by nuclear fission to different chemical processes. The KFA is involved in the two German projects PNP - Prototypanlage Nukleare Prozesswaerme (Prototype-plant Nuclear Process-heat), and NFE -Nukleare Fernenergie (Long Distance Energy Transport). In a HTGR, helium generally serves as reactor coolant. It transports the heat from the core to the different components which take over this heat for various purposes. In case of arranging a steam reformer in the helium circuit, it is necessary for economic reasons to reach very high temperatures. In the two German projects mentioned above, the helium temperature at HTGR core outlet is determined to 9500C. Thus the main design data for a steam reformer supplied by heat from a HTGR are maximum helium temperature 9500C, helium pressure 40 bar. By an extensive utilization of the available advanced conventional steam reforming technology, the helium heated steam reformer design is using normal steam reforming tubes arranged in compact bundles

  20. Neutron spectrum measurements from a neutron guide tube facility at the ETRR-1 reactor

    International Nuclear Information System (INIS)

    The present work deals with measurements of the neutron spectrum emitted from a neutron guide tube (NGT) recently installed at one of the ETRR-1 reactor horizontal channels designed to deliver thermal neutrons, free from fast neutrons and gamma ray background, to a fourier reverse-time-of-flight (RTOF) diffractometer. The measurements were performed using a 6 Li glass scintillation detector combined with a multichannel analyzer set at channel width 4 M sec and installed at 3.4 m from a disc Fermi chopper. Also a theoretical model was specially developed for the neutron spectrum calculations. According to the model developed, the spectrum calculated was found to be in good agreement with the measured one. It was found, both from measurements and calculations, that the spectrum emitted from the NGT covers, after transmission through a fourier chopper, neutron wavelengths from 1-4 A adequate for neutron diffraction measurements at D values between 0.71-2.9 A respectively. 6 FIGS

  1. Neutron Spectrum Measurements from a Neutron Guide Tube Facility at the ETRR-1 Reactor

    International Nuclear Information System (INIS)

    The present work deals with measurements of the neutron spectrum emitted from a neutron guide tube (NGT) recently installed at one of the ETRR-1 reactor horizontal channels; designed to deliver thermal neutrons, free from fast neutrons and gamma-rays background, to a Fourier reverse-time-of-flight (RTOF) diffractometer. The measurements were preformed using a 6Li glass scintillation detector installed at 3.45 m from a disc Fermi chopper, combined with a multichannel analyzer set at channel width 4 mu sec. Also a theoretical model was especially developed for the neutron spectrum calculations. The spectrum calculated according to the developed model was found to be in good agreement with the measured one. It was found, both from measurements and calculations, that the spectrum emitted from the NGT covers, after transmission through a Fourier chopper, neutron wavelengths from 1-4 A degree. The maximum of the neutron spectrum was found to be at neutron wavelength lambda=1.39 A degree which is consistent with the value 1.377 A degree calculated for the curved NGT characteristic wavelength

  2. Influence of the biogas reburning for reducing nitric oxide emissions in an alundum-tube reactor

    Science.gov (United States)

    Zhao, Jie; Wang, Qingcheng; Yu, Lihui; Wu, Liyan

    2016-05-01

    The experimental study on reburning reduction reaction between biogas and NO is very important in de-NOx technology. The reburning experiments by the simulated biogas with different operation variables have been performed in an alundum-tube reactor. Results showed that the uppermost constituent in NO-reduction was CH4, H2 second, and NO-reduction by CO in biogas reburning was negligible at the same conditions. In the condition of oxygen-poor, H2 could promote CH4 oxidation and enhance the concentration of CH3 radicals, thereby increasing the reduction efficiency of NO accordingly. At the same temperature, with the increase of stoichiometric ratio, it would increase O radicals and decrease NO reduction efficiency. With the increase of reaction temperature, the reduction efficiency behaved a trend of first increased then decreased at the same stoichiometric ratio, and obtained the maximum value 51.38% at the condition of 1200 °C and λ = 0.6. Additionally, increasing the NO input concentration also could improve the reduction efficiency under the condition of fuel-rich.

  3. Influence of the biogas reburning for reducing nitric oxide emissions in an alundum-tube reactor

    Science.gov (United States)

    Zhao, Jie; Wang, Qingcheng; Yu, Lihui; Wu, Liyan

    2016-05-01

    The experimental study on reburning reduction reaction between biogas and NO is very important in de-NOx technology. The reburning experiments by the simulated biogas with different operation variables have been performed in an alundum-tube reactor. Results showed that the uppermost constituent in NO-reduction was CH4, H2 second, and NO-reduction by CO in biogas reburning was negligible at the same conditions. In the condition of oxygen-poor, H2 could promote CH4 oxidation and enhance the concentration of CH3 radicals, thereby increasing the reduction efficiency of NO accordingly. At the same temperature, with the increase of stoichiometric ratio, it would increase O radicals and decrease NO reduction efficiency. With the increase of reaction temperature, the reduction efficiency behaved a trend of first increased then decreased at the same stoichiometric ratio, and obtained the maximum value 51.38% at the condition of 1200 °C and λ = 0.6. Additionally, increasing the NO input concentration also could improve the reduction efficiency under the condition of fuel-rich.

  4. Chirp trackline navigation data from USGS cruise 2003-042-FA from Pamlico Sound, North Carolina (bbc2003042_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  5. Chirp navigation tracklines from USGS cruise 2004-003-FA along the inner continental shelf of northern North Carolina (isc2004003_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  6. Chirp shotpoint navigation from USGS cruise 2002-012-FA along the inner continental shelf of northern North Carolina (isc2002012_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  7. Chirp shotpoint navigation from USGS cruise 2001-005-FA along the inner continental shelf of northern North Carolina (isc2001005_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  8. Sidescan sonar navigation from USGS cruise 1999-045-FA along the inner continental shelf of northern North Carolina (iss1999045_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  9. Sidescan sonar navigation from USGS cruise 2003-003-FA along the inner continental shelf of northern North Carolina (iss2003003_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  10. Sidescan sonar navigation from USGS cruise 2002-013-FA along the inner continental shelf of northern North Carolina (iss2002013_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  11. Chirp navigation tracklines from USGS cruise 2003-003-FA along the inner continental shelf of northern North Carolina (isc2003003_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  12. Boomer shotpoint navigation from USGS cruise 2002-012-FA along the inner continental shelf of northern North Carolina (isb2002012_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  13. Chirp navigation tracklines from USGS cruise 2001-005-FA along the inner continental shelf of northern North Carolina (isc2001005_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  14. Sidescan sonar navigation from USGS cruise 2002-012-FA along the inner continental shelf of northern North Carolina (iss2002012_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  15. isc2002012_tracklines.shp: Chirp navigation tracklines from USGS cruise 2002-012-FA along the inner continental shelf of northern North Carolina

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  16. Chirp shotpoint navigation from USGS cruise 2002-013-FA along the inner continental shelf of northern North Carolina (isc2002013_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  17. Boomer shotpoint navigation from USGS cruise 2001-005-FA along the inner continental shelf of northern North Carolina (isb2001005_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  18. Boomer shotpoint navigation from USGS cruise 1999-045-FA along the inner continental shelf of northern North Carolina (isb1999045_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  19. Boomer seismic trackline data from USGS cruise 2001-013-FA from Albemarle Sound, North Carolina (bbb2001013_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  20. Boomer seismic trackline data from USGS cruise 2001-005-FA along the inner continental shelf of northern North Carolina (isb2001005_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  1. Boomer seismic trackline data from USGS cruise 2002-012-FA along the inner continental shelf of northern North Carolina (isb2002012_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  2. Boomer shotpoint navigation from USGS cruise 2002-013-FA along the inner continental shelf of northern North Carolina (isb2002013_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  3. Chirp navigation tracklines from USGS cruise 2002-013-FA along the inner continental shelf of northern North Carolina (isc2002013_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  4. Boomer seismic trackline data from USGS cruise 2002-013-FA along the inner continental shelf of northern North Carolina (isb2002013_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  5. Chirp shotpoint navigation from USGS cruise 1999-045-FA along the inner continental shelf of northern North Carolina (isc1999045_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  6. Sidescan sonar navigation from USGS cruise 2004-003-FA along the inner continental shelf of northern North Carolina (iss2004003_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  7. Boomer seismic trackline data from USGS cruise 1999-045-FA along the inner continental shelf of northern North Carolina (isb1999045_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  8. Chirp shotpoint navigation from USGS cruise 2003-003-FA along the inner continental shelf of northern North Carolina (isc2003003_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  9. Chirp navigation tracklines from USGS cruise 1999-045-FA along the inner continental shelf of northern North Carolina (isc1999045_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  10. Chirp trackline navigation data from USGS cruise 2003-005-FA from Pamlico Sound, North Carolina (bbc2003005_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  11. Sidescan sonar navigation from USGS cruise 2001-005-FA along the inner continental shelf of northern North Carolina (iss2001005_tracklines.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  12. Location of MMS (Minerals Management Service) vibracores from offshore northern Dare County, North Carolina (mms_cores.shp, geographic, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  13. Location of SNL vibracores collected on Debris Barge (D/B) Snell from offshore northern Dare and Hyde Counties, North Carolina (snl_cores.shp, geographic, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  14. Chirp shotpoint navigation from USGS cruise 2004-003-FA along the inner continental shelf of northern North Carolina (isc2004003_shots.shp)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  15. Experimental determination of thermal contact conductance between pressure and calandria tubes of Indian pressurised heavy water reactors

    International Nuclear Information System (INIS)

    Highlights: • We established an experimental facility to measure thermal contact conductance between disc shaped specimens. • We measured thermal contact conductance between Zr-2.5Nb alloy pressure tube (PT) material and Zr-4 calandria tube (CT) material. • We concluded that thermal contact conductance is a linear function of contact pressure for interface of PT and CT up to 10 MPa contact pressure. • We concluded that thermal contact conductance is a weak function of interface temperature. - Abstract: Thermal contact conductance (TCC) is one of the most important parameters in determining the temperature distribution in contacting structures. Thermal contact conductance between the contacting structures depends on the mechanical properties of underlying materials, thermo-physical properties of the interstitial fluid and surface condition of the structures coming in contact. During a postulated accident scenario of loss of coolant with coincident loss of emergency core cooling system in a tube type heavy water nuclear reactor, the pressure tube is expected to sag/balloon and come in contact with outer cooler calandria tube to dissipate away the heat generated to the moderator. The amount of heat thus transferred is a function of thermal contact conductance and the nature of contact between the two tubes. An experimental facility was designed, fabricated and commissioned to measure thermal contact conductance between pressure tube and calandria tube specimens. Experiments were conducted on disc shaped specimens under axial contact pressure in between mandrels. Experimental results of TCC and a linear correlation as a function of contact pressure have been reported in this paper

  16. Experimental determination of thermal contact conductance between pressure and calandria tubes of Indian pressurised heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dureja, A.K., E-mail: akdureja@barc.gov.in [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Pawaskar, D.N.; Seshu, P. [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Mumbai (India); Sinha, S.K. [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Sinha, R.K. [Department of Atomic Energy, OYC, Near Gateway of India, Mumbai (India)

    2015-04-01

    Highlights: • We established an experimental facility to measure thermal contact conductance between disc shaped specimens. • We measured thermal contact conductance between Zr-2.5Nb alloy pressure tube (PT) material and Zr-4 calandria tube (CT) material. • We concluded that thermal contact conductance is a linear function of contact pressure for interface of PT and CT up to 10 MPa contact pressure. • We concluded that thermal contact conductance is a weak function of interface temperature. - Abstract: Thermal contact conductance (TCC) is one of the most important parameters in determining the temperature distribution in contacting structures. Thermal contact conductance between the contacting structures depends on the mechanical properties of underlying materials, thermo-physical properties of the interstitial fluid and surface condition of the structures coming in contact. During a postulated accident scenario of loss of coolant with coincident loss of emergency core cooling system in a tube type heavy water nuclear reactor, the pressure tube is expected to sag/balloon and come in contact with outer cooler calandria tube to dissipate away the heat generated to the moderator. The amount of heat thus transferred is a function of thermal contact conductance and the nature of contact between the two tubes. An experimental facility was designed, fabricated and commissioned to measure thermal contact conductance between pressure tube and calandria tube specimens. Experiments were conducted on disc shaped specimens under axial contact pressure in between mandrels. Experimental results of TCC and a linear correlation as a function of contact pressure have been reported in this paper.

  17. Nondestructive inspection of the tubes of TRIGA IPR-R1 reactor heat exchanger by eddy current testing

    Energy Technology Data Exchange (ETDEWEB)

    Silva Junior, Silverio F.; Silva, Roger F.; Oliveira, Paulo F., E-mail: silvasf@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Barreto, Erika S.; Ribeiro, Isabela G.; Fraiz, Felipe C. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2013-07-01

    The IPR-R1 TRIGA MARK 1 reactor is an open pool type reactor, cooled light water. It is used for research activities, personnel training and radioisotopes production, in operation since 1960 at the Nuclear Technology Development Center - CDTN/CNEN. It operates at a maximum thermal power of 100 kW and usually, the fuel cooling is done by natural circulation. If necessary, an external auxiliary cooling system, with a shell-and-tube type heat exchanger, can be used to improve the water heat removal. As part of the ageing management program of the reactor, a nondestructive evaluation of their heat exchanger stainless steel tubes will be performed, in order to verify its integrity. The examinations will be performed using the eddy current test method, which allows the detection and characterization of structural discontinuities in the wall of the tubes, if existing. For this purpose, probes and reference standards were designed and manufactured at CDTN facilities and test procedures were established and validated. In this paper, a description of the proposed infrastructure as well as the test methodology to be used in the examinations are presented and discussed. (author)

  18. Heat transfer characteristics of sodium-water reaction jet around a tube in steam generator of sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Overheating tube rupture of adjacent tubes arises from water/steam leak in steam generators of sodium-cooled fast reactors. It is very important to predict the tube wall stress (tube wall temperature) with a high degree of accuracy on evaluation of overheating tube rupture, and is crucial to estimate quantitatively the heat transfer coefficient between reaction jet and adjacent tubes which is one of the major influencing factor. The authors carried out the sodium-water reaction test (SWAT-1R) under the simulated operation condition of a real plant, and measured the correlation between heat transfer coefficient and void fraction around an adjacent tube. The authors confirmed that thermal environment around an adjacent tube was inferable from measured data, and heat transfer correlation equation proposed by Hamada et al. was applicable to the operation condition at elevated pressure and temperature. (author)

  19. Serpentine tube heat transfer characteristic under accident condition in gas cooled reactors

    International Nuclear Information System (INIS)

    In nuclear reactors of the Magnox or advanced gas Cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accident conditions using two phase flow codes requires knowledge of the heat transfer behavior of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear Electric. The tests were carried out on the Thermal Hydraulics Experimental Research Assembly (THERA) loop at Manchester University. The Thermal Hydraulic Experimental Research Assembly was designed to operate with pressures up to 180 bar and temperatures of 450degC. The geometry and dimensions of this test section were similar to part of a gas cooled reactor boiler of the Hinkley Point design. Blowdown from a pressure of 60 bar with subcoolings of 5degC, 50degC, 100degC formed the main part of the programme. A set of tests was conducted using discharge orifices of different sizes to produce depressurization times from 30 s to 10 mins, and in a few cases, the duration of blowdown approached 1 hour. These times were defined using the criterion of blowdown end as a final pressure of 10% of the initial pressure. Pressures, wall and fluid temperatures were all measured at average time intervals of 1.1s during the excursion and an inventory of the remaining water content in the serpentine was taken when the blowdown ended. Some tests were also conducted at an initial pressure of 30 bar. The results obtained show interesting stratification effects for the relatively fast discharge, with substantial wall circumferential temperature variations. For these tests, a relatively small water inventory remained after blowdown. The discharge characteristics of the serpentine in terms of orifice size have been mapped, and tests at 30 bar show the equivalence in terms of orifice size have been mapped

  20. Study of the U/sub 3/O/sub 8/-Al thermite reaction and strength of reactor fuel tubes

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, H.B.

    1983-08-01

    Heating tests using 53 wt % U/sub 3/O/sub 8/-Al pellets show that an exothermic reaction occurs between 875 and 1000/sup 0/C and takes 10 to 20 seconds to reach maximum temperature. The maximum temperature is a function of particle size of the U/sub 3/O/sub 8/ with large particles exhibiting lower peak temperatures. The calculated energy release was 123 cal/g of U/sub 3/O/sub 8/-aluminum fuel. Tests using aluminum clad outer fuel tube sections gave lower peak temperatures than for pellets. No violent reactions occurred. The results are reasonably consistent with recent reported data indicating that the exothermic U/sub 3/O/sub 8/-Al reaction is not an important energy source. The compressive and tensile strengths of U/sub 3/O/sub 8/ tubes above 660/sup 0/C are low. In compression, sections with 2 psi average axial stress failed at 917/sup 0/C, while sections with 7 psi failed at 669/sup 0/C. Tubes with U-Al alloy cores failed at about 670/sup 0/C with no applied load. The stresses in fuel tubes during a reactor transient may range up to several hundred psi and are less than 7 psi only in the upper part of the fuel tube.

  1. Natural gas pyrolysis in double-walled reactor tubes using thermal plasma or concentrated solar radiation as external heating source

    Institute of Scientific and Technical Information of China (English)

    Stèphane Abanades; Stefania Tescari; Sylvain Rodat; Gilles Flamant

    2009-01-01

    The thermal pyrolysis of natural gas as a clean hydrogen production route is examined.The concept of a double-walled reactor tube is proposed and implemented.Preliminary experiments using an external plasma heating source are carded out to validate this concept.The results point out the efficient CH4 dissociation above 1850 K (CH4 conversion over 90%) and the key influence of the gas residence time.Simulations are performed to predict the conversion rate of CH4 at the reactor outlet,and are consistent with experimental tendencies.A solar reactor prototype featuring four independent double-walled tubes is then developed.The heat in high temperature process required for the endothermic reaction of natural gas pyrolysis is supplied by concentrated solar energy.The tubes are heated uniformly by radiation using the blackbody effect of a cavity-receiver absorbing the concentrated solar irradiation through a quartz window.The gas composition at the reactor outlet,the chemical conversion of CH4,and the yield to H2 are determined with respect to reaction temperature,inlet gas flow-rates,and feed gas composition.The longer the gas residence time,the higher the CH4 conversion and H2 yield,whereas the lower the amount of acetylene.A CH4 conversion of 99% and H2 yield of about 85% are measured at 1880 K with 30% CH4 in the feed gas (6 L/min injected and residence time of 18 ms).A temperature increase from 1870 K to 1970 K does not improve the H2 yield.

  2. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes. Determination of hydrogen concentration and blister characterization

    International Nuclear Information System (INIS)

    Heavy water reactors (HWRs) comprise significant numbers of today's operating nuclear power plants, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes, are an important factor in ensuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Project (CRP) on Intercomparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the framework of the IAEA's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of IAEA's project on advanced technologies for HWRs. The objective of the CRP was to compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP participants investigated the capability of different techniques to detect and characterize flaws. During the second phase participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in zirconium alloys. The intention was to identify the most effective pressure tube inspection and diagnostic methods and to identify further development needs. The organizations which participated in phase 2 of this CRP are: - Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL), Chalk River Laboratories (CRL), Canada; - Bhabha Atomic Research Centre (BARC), India; - Korea Atomic Energy Research Institute (KAERI), Republic of Korea; - National Institute for Research and Development for Technical Physics (NIRDTP), Romania; - Nuclear Non-Destructive Testing Research and Services (NNDT), Romania. IAEA-TECDOC-1499

  3. The second eddy current testing of zircaloy tube samples from the OECD Halden reactor project at Reactor Fuel Examination Facility, Tokai, JAERI

    International Nuclear Information System (INIS)

    The Reactor Fuel Examination Facility in Tokai/JAERI (Japan Atomic Energy Research Institute) joined to the second round robin programme on eddy current test of the Halden/IFE. In the programme, two zircaloy tube samples with some artificial defects were provided for measurements. To clarify the locations in axial and azimuthal directions, types and dimensions of the provided artificial defects, measured signals from eddy current test were analysed in comparison with the known defects on the calibration tube. As a result, fourteen defects were determined from the measurements. Then, the location, the type and the relative dimension of them were also revealed. The results of those eddy current test are described in this paper. (author)

  4. Application of automatic inspection system to nondestructive test of heat transfer tubes of primary pressurized water cooler in the high temperature engineering test reactor. Joint research

    International Nuclear Information System (INIS)

    Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the heat transfer tube of the PPWC, were different from those obtained for the artificially defected tube. In UT echo amplitude of the PPWC tubes inspected was lower than 20% of distance-amplitude calibration curve. Thus, it was confirmed that there was no defect in depth, which was more than the detecting standard of the probes, on the outer surface of the heat transfer tubes of the PPWC inspected. (author)

  5. Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

    2014-06-15

    Highlights: • Detailed burnup calculations are performed on pressurized water cooled blankets with pressure tube assemblies. • The blanket is fueled with simple fuel, namely spent nuclear fuel discharged from light water reactors or natural uranium oxide. • The refueling strategies are proposed, and the uranium resource utilization rate can reach 5–6%. - Abstract: A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.

  6. Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Highlights: • Detailed burnup calculations are performed on pressurized water cooled blankets with pressure tube assemblies. • The blanket is fueled with simple fuel, namely spent nuclear fuel discharged from light water reactors or natural uranium oxide. • The refueling strategies are proposed, and the uranium resource utilization rate can reach 5–6%. - Abstract: A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%

  7. Production and control of stainless steel cladding tubes for breeder reactors

    International Nuclear Information System (INIS)

    The reliability of a cladding tube in a LMFBR depends, for the most part, on its capability of keeping its original shape, even under extremely heavy duty conditions. A tremendous effort has been conducted in two different ways, to attain this purpose: definition of the metallurgical characteristics of the cladding tube material according to its mechanical properties in high temperature conditions and its properties under neutronic action; an improved quality of the tubes by methodical and controlled processing conducted at Vallourec'plant in Montbard in a specially developed workshop for this type of tubes. The raw material controls, the tubes production scheduling in hot and cold conditions, together with the non-destructive and destructive tests are defined both by Vallourec and the C.E.A

  8. High-conversion and high-burnup core concepts for pressure-tube-type heavy water reactors

    International Nuclear Information System (INIS)

    A high-conversion and a high-burnup core concept for a pressure-tube-type heavy water reactor are presented and analyzed from the standpoint of neutronics. These core concepts are based on the fact that neutron spectrum can be shifted by adjusting the amount of heavy water moderator outside the pressure tubes without affecting core-cooling capability. For the high-conversion core, where the heavy water moderator is replaced by a gas such as CO2 [carbon dioxide], a conversion ratio of more than 0.8 and an average discharge fuel burnup of 50GWd/t have been estimated to be attained with standard design fuel assemblies having 7.5% fissile Pu enrichment. For the high-burnup core, where fuel assemblies burned in the high-conversion (gas) region are relocated into the burner (heavy water) region, an average discharge fuel burnup of 110GWd/t has been estimated

  9. Control Rod Ejection Accident while Using 6- and 8-Tube IRT-4M Fuel Assemblies in WWR-SM Research Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Baytelesov, S.; Kungurov, F.; Safarov, A.; Salikhbaev, U.

    2011-07-01

    The WWR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences (INP AS) in Uzbekistan was converted to 6-tube IRT-4M LEU (19.7%) fuel in 2009. Presently, INP intends to also use IRT-4M 8-tube FA, and a safety analysis for these 'mixed' (8-tube and 6-tube FA) cores is required by the regulatory authorities. This paper presents results of control rod ejection transient analysis for these mixed cores

  10. 4 meter sidescan-sonar GeoTIFF image of inner shelf from Virginia border to Cape Hatteras, NC (composite_nhatt.tif, UTM, Zone 18N, WGS84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  11. 4 meter sidescan-sonar GeoTIFF image of inner shelf with stretched histogram, from Virginia border to Cape Hatteras, NC (composite_nhatt_str.tif, UTM, Zone 18N, WGS84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  12. 40 meter ESRI binary grid of single beam and swath bathymetry of inner continental shelf north of Cape Hatteras, NC to Virginia border (nhatt, UTM Zone 18N, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  13. Corrosion study of heat exchanger tubes in pressurized water cooled nuclear reactors by conversion electron Moessbauer spectroscopy

    International Nuclear Information System (INIS)

    Nuclear energy production tends to return into the focus of interest because of the constantly increasing energy need of the world and the green house effect problems of the strongest competitor oil or gas based power plants. In addition to the construction of new nuclear power plants, lifetime extension of the existing ones is the most cost effective investment in the energy business. However, feasibility and safety issues become very important at this point, and corrosion of the construction materials should be carefully investigated before decision on a potential lifetime extension of a reactor. 57Fe-Conversion Electron Moessbauer Spectroscopy (CEMS) is a sensitive tool to analyze the phase composition of corrosion products on the surface of stainless steel. The upper ∼300 nm can be investigated due to the penetration range of conversion electrons. The corrosion state of heat exchanger tubes from the four reactor units of the Paks Nuclear Power Plant, Hungary, were analyzed by several methods including CEMS. The primary circuit side of the tubes was studied on selected samples cut out from the heat exchangers during regular maintenance. Cr- and Ni-substituted magnetite, sometimes hematite, amorphous Fe-oxides/oxyhydroxides as well as the signal of bulk austenitic steel of the tubes were detected. The level of Cr- and Ni-substitution in the magnetite phase could be estimated from the Moessbauer spectra. Correlation between earlier decontamination cycles and the corrosion state of the heat exchangers was sought. In combination with other methods, a hybrid structure of the surface oxide layer of several microns was established. It is suggested that previous AP-CITROX decontamination cycles can be responsible for this structure which makes the oxide layer mobile. This mobility may be responsible for unwanted corrosion product transport into the reactor vessel by the primary coolant.

  14. In-service inspection of zircaloy pressure tube of CIRUS reactor

    International Nuclear Information System (INIS)

    The pressurized water loop (PWL) of Cirus uses a 10 meter long zircaloy tube of 57.9 mm ID and 5.4 mm wall thickness. The loop has been used for irradiation testing of various experimental fuel pins since 1972. As part of the refurbishment programme, the condition of Zircaloy-2 pressure tube of the pressurized water loop was investigated by Eddy current and ultrasonic testing. The eddy current probe was balanced over a portion of the tube and the differential signals were recorded for the entire length of the tube. For ultrasonic flaw scanning, gadgets were fabricated and scanning was carried out to evaluate the condition of irradiated pressure tube. For ultrasonic testing an annular probe holder matching to the internal diameter of the zircaloy tube was used for immersion scanning. The probe holder fitted with 10 MHz line focused ultrasonic probes inclined at 28 deg in axial and circumferential directions. A normal spot focused probe was also used to measure wall thickness and detection of laminar flaws. Axial and circumferential grooves of 3% wall thickness depth on ID and OD were used as standard calibration defects. The eddy current and ultrasonic tests did not detect any defect of unacceptable size in the zircaloy pressure tube. (author)

  15. Control rod drive/reactor vessel stub tube penetration experience and repair programmes

    International Nuclear Information System (INIS)

    Two heats of furnace sensitized Type 304 stainless steel were used in the material qualification tests for the Nuclear stub tube repair program. The carbon contents of the steels were 0.047 and 0.06 percent, which were intentionally selected to be conservatively higher than the actual stub tube materials. The furnace sensitized treatment consisted of heating at 1150 F (621 C) for 20 hours followed by furnace cooling. Welding was performed on the steel after furnace sensitizing to simulate the welding required to install a partial sleeve onto the stub tube. (author)

  16. Analysis of the Hydrogen Reduction Rate of Magnetite Concentrate Particles in a Drop Tube Reactor Through CFD Modeling

    Science.gov (United States)

    Fan, Deqiu; Mohassab, Yousef; Elzohiery, Mohamed; Sohn, H. Y.

    2016-06-01

    A computational fluid dynamics (CFD) approach, coupled with experimental results, was developed to accurately evaluate the kinetic parameters of iron oxide particle reduction. Hydrogen reduction of magnetite concentrate particles was used as a sample case. A detailed evaluation of the particle residence time and temperature profile inside the reactor is presented. This approach eliminates the errors associated with assumptions like constant particle temperature and velocity while the particles travel down a drop tube reactor. The gas phase was treated as a continuum in the Eulerian frame of reference, and the particles are tracked using a Lagrangian approach in which the trajectory and velocity are determined by integrating the equation of particle motion. In addition, a heat balance on the particle that relates the particle temperature to convection and radiation was also applied. An iterative algorithm that numerically solves the governing coupled ordinary differential equations was developed to determine the pre-exponential factor and activation energy that best fit the experimental data.

  17. Location of rotasonic cores from northeastern North Carolina Volumes I to IV: Cores OBX-01 through OBX-18 and MLD-01 through MLD-10 (obx_cores.shp, geographic, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  18. Analytical study of nuclear-coupled density-wave instability in a natural circulation pressure tube type boiling water reactor

    International Nuclear Information System (INIS)

    An analytical model has been developed to study the nuclear-coupled density-wave instability in the Indian advanced heavy water reactor (AHWR) which is a natural circulation pressure tube type boiling water reactor. The model considers a point kinetics model for the neutron dynamics and a lumped parameter model for the fuel thermal dynamics along with the conservation equations of mass, momentum and energy and equation of state for the coolant. In addition, to study the effect of neutron interactions between different parts of the core, the model considers a coupled multipoint kinetics equation in place of simple point kinetics equation. Linear stability theory was applied to reveal the instability of in-phase and out-of-phase modes in the boiling channels of the AHWR. The results indicate that the stability behavior of the reactor is greatly influenced by the void reactivity coefficient, fuel time constant, radial power distribution and channel inlet orificing. The delayed neutrons were found to have a strong influence on the Type I and Type II instabilities observed at low and high channel powers, respectively. Also, it was found that the coupled multipoint kinetics model and the modal point kinetics model predict the same threshold power for out-of-phase instability if the coupling coefficient in the former model is half the eigen value separation between the fundamental and the first harmonic mode in the latter model. Decay ratio maps were predicted considering various operating parameters of the reactor, which are useful for its design. (orig.)

  19. Multiple pressure tube rupture in channel type reactors. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    Accident scenarious which could potentially lead to multiple pressure tube ruptures have been recognized as a major safety issue for the RBMK. Therefore, a topical meeting on multiple pressure tube rupture analysis in channel type reactors was convened by the IAEA at the RDIPE in Moscow from 31 January to 4 February 1994 in the framework of the IAEA Extrabudgetary Programme on the Safety of RBMK Nuclear Power Plants. The objective of the meeting was to exchange experience on approaches adopted in Member States operating channel type reactors and review analysis methodology, criteria and results obtained related to multiple pressure tube rupture in RBMK type reactors including related regulatory requirements. The review was carried out in two broader technical areas: Pressure tube integrity and potential for failure propagation; Multiple pressure tube rupture scenarios. The following conclusions have been derived: Propagation of a single tube rupture is unlikely; The analysis of accident scenarios which could lead to multiple pressure tube ruptures in RBMKs is among the highest priority safety issues. International co-operation including independent analyses by experts from OECD countries, is required to resolve this issue; Further experimental information is needed both to validate the computer codes and to obtain a better understanding of some of the physical phenomena involved; No specific scenario for multiple failures following a localized power excursion was identified. 2 refs, 12 figs, 4 tabs

  20. Study of the U/sub 3/O/sub 8/-Al thermite reaction and strength of reactor fuel tubes

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, H B

    1983-01-01

    Research and test reactors are presently operated with aluminum-clad fuel elements containing highly enriched uranium-aluminum alloy cores. To lower the enrichment and still maintain reactivity, the uranium content of the fuel element will need to be higher than currently achievable with alloy fuels. This will necessitate conversion to other forms such as U/sub 3/O/sub 8/-aluminum cermets. Above the aluminum melting point, U/sub 3/O/sub 8/ and aluminum undergo an exothermic thermite reaction and cermet fuel cores tend to keep their original shape. Both factors could affect the course and consequences of a reactor accident, and prompted an investigation of the behavior of cermet fuels at elevated temperatures. Tests were carried out using pellets and extruded tube-sections with 53 wt % U/sub 3/O/sub 8/ in aluminum. This content corresponds to a theoretical uranium density of 1.9 g/cc. Results indicate that the thermite reaction occurs at about 900/sup 0/C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U/sub 3/O/sub 8/-aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 660/sup 0/C. In tension, sections failed at about the aluminum melting point. In compression with 2-psi average axial stress, failure occurred at 917/sup 0/C, while 7 psi average axial stress produced failure at 669/sup 0/C.

  1. Advanced NDE (ANDE) and its application for pressure tube inspections in OPG reactors

    International Nuclear Information System (INIS)

    Periodic and in-service inspections of CANDU fuel channels are essential for the proper assessment of the structural integrity of these vital components. The arrival of new delivery devices for fuel channel inspections (Universal Delivery Machine) has driven new methods for gathering and analyzing NDE data. The Advanced Non-Destructive Examination (ANDE) system has been designed and field implemented as a high speed data acquisition system to meet the requirements of the CSA N285.4 code. It was built from the solid foundation of CIGAR experience and uses cutting edge hardware and software to attain high speed data collection enabling relatively quick inspection of a large number of fuel channels. The capabilities of the ANDE inspection system include: Surface and volumetric inspection of pressure tube by ultrasonics; Flaw characterization by ultrasonics; Pressure tube diameter measurements; Pressure tube thickness measurements; Garter Spring location by Eddy Current; Garter Spring location by ultrasonics; Pressure tube sag measurement. In addition to the above, selected flaws/areas of a pressure tube can be replicated using a two plate ANDE replica tool. At the heart of the inspection system is a set of twelve ultrasonic probes positioned in such a way that the inspected areas are examined from various angles and directions and by various ultrasonic wave modes (shear and longitudinal). High frequency ultrasound used for the examinations allows for reliable detection of small flaws. Separate sensors have been installed on the inspection head for Garter Spring location and sag measurements. (author)

  2. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  3. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    International Nuclear Information System (INIS)

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days

  4. Ageing Management and Preventice Measures for Reactor Pool Liners, Beam Tubes and Spent Fuel Storage Tank at the Dalat Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dien, Nguyen Nhi; Dien, Nguyen Minh; Su, Trang Cao [Nuclear Research Institute, Henoi (Viet Nam)

    2013-07-01

    The 500-kw Dalat Nuclear Research Reactor (DNRR) was reconstructed from the original 250-kW TRIGA Mark II as named of VN-001. In the framework of the reconstruction project during the 1982-1984 period, some structures of the TRIGA reactor constructed in the early sixties, such as the aluminum tank, graphite reflector, thermal column, four horizontal beam tubes, etc. have been remained. It means, such components are more than 50 years old and are facing with ageing issues. The structural materials of the pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. The reactor core has been loaded with HEU VVR-M2 fuel assemblies of 36% enrichment alloy CAV-1. The reactor core has been loaded with HEU VVR-M2 fuel assemblies of U-Al alloy 36% and of UO{sub 2} 19.75% enrichment used aluminum as fuel cladding. For ageing management and preventive measures of corrosion, an underwater high-resolution video camera system had been designed for visual inspections. A home-made cleaning system was also designed for cleaning the pool and other components. Water chemistry of the reactor pool and spent fuel storage was monitored regularly. In September-November 2011, all four horizontal channels were cleaned inside and visual inspection was done using special camera system. It was the first time from 1963 such activity could be done. Based on results obtained we could convince that inside all horizontal channels are in good condition and leakage could not be occurred. All 106 HEU spent fuel assemblies stored in the spent fuel pool in good condition. The visual inspection was done using under water camera too. The results obtained show that the surface of all HEU SFA is good and leakage was not occurred. The

  5. Ultrasonic testing of canning tubes in stainless steel of the EL 4 reactor

    International Nuclear Information System (INIS)

    From all the methods possible for controlling thin cans the one chosen, for numerous reasons, vas that making use of ultrasonic techniques. A method has been developed which should make it possible to carry out a rapid and efficient industrial control of canning tubes, The reasons for the choice of the ultrasonic method are given in detail, together with the principles of the method and the actual control parameters. In the present state of our research, it should be possible to control at least 50 000 tubes a year. Improvements brought about in the details of the control technique itself should make it possible to increase this rate considerably. (authors)

  6. Application of RCC-MR for the structural design of tube sheet of intermediate heat exchanger for a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Highlights: • The structural mechanics behavior of tube sheets has been studied in detailed by FEM analysis software CAST3M. • Practical guidelines are provided to accurately derive the primary stress intensities particularly at the interface. • Linearization of stress components along the radial directions and the effect of filet radius at the interfaces have been covered in detailed. • Estimation of creep–fatigue damage as per RCC-MR. • These investigations thus help to optimize the design of IHX tube sheets with high confidence. - Abstract: The structural mechanics behavior of tube sheets of a sodium to sodium heat exchanger for a fast reactor, with circular tube holes pattern, less addressed subject in the literature, is investigated in detail. The tube sheet design rules recommended in the French design code RCC-MR-2007 and the associated solid mechanics basis are explained. A finite element analysis of tube sheets of intermediate heat exchanger of a typical 500 MWe pool type fast reactor is presented to study the effects of some specific parameters viz., (i) small solid rim portion with connecting shell and (ii) grooves on rim area. For the analysis, the distribution of holes on the last row is assumed to be symmetric and axial stiffening of tubes on tube sheet is included toward realistic estimation of stresses in the tube sheets. The effects are studied on the primary and secondary stresses induced along the interface between solid to perforated region. The aspects covered include linearization of radial and circumferential stress components, thereby deriving primary membrane and bending stress intensities along the radial directions with particular focus at the interfaces between solid portions and perforated portions including the effect of filet radius at the junction of tube sheet and shell. These investigations thus help to optimize the design of IHX tube sheets with high confidence. The analysis has been carried out by CAST3M, a structure

  7. Virginia Tech Corps of Cadets alumnus Thomas Lenz named Hokie Hero

    OpenAIRE

    Cox, Carrie

    2009-01-01

    Virginia Tech Corps of Cadets alumnus 2nd Lt. Thomas Lenz, United States Army, who earned a degree in history from the College of Liberal Arts and Human Sciences in 2008 has been selected as the Hokie Hero for the Virginia Tech versus North Carolina State football game

  8. Virginia Tech's Corps of Cadets alumnus Brian Alberts named Hokie Hero

    OpenAIRE

    Cox, Carrie

    2009-01-01

    Virginia Tech Corps of Cadets alumnus 2nd Lt. Brian Alberts, United States Army, who earned a degree in history from the College of Liberal Arts and Human Sciences in 2008 has been selected as the Hokie Hero for the Virginia Tech versus East Carolina University football game.

  9. Artificial Neural Network Modelling of In-Reactor Diametral Creep of Zr2.5%Nb Pressure Tubes of Indian PHWRs

    International Nuclear Information System (INIS)

    Highlights: • An ANN model is developed to predict the in-reactor diametral creep in Zr–2.5%Nb. • Database from pressure tubes of Indian PHWRs have been used. • The developed ANN model can efficiently predict diametral creep of pressure tubes. • O cont and mech properties play important role in determining diametral creep rate. - Abstract: A model is developed to predict the in-reactor diametral creep in the Zr–2.5%Nb pressure tube of Indian Pressurized Heavy Water power reactors (PHWR) using Artificial Neural Network (ANN). The inputs of the neural network are alloy composition of the tube (concentration of Nb, O, N and Fe), mechanical properties (YS, UTS, %EL), temperature and fluence whereas diametral creep rate is the output. Measured diametral creep rate data from the sampled pressure tubes operating in Indian PHWRs at Rajasthan Atomic Power Station (RAPS 2), Kakrapar Atomic Power Station (KAPS 2) and Kaiga Generating Station (KGS) are employed to develop the model. A three-layer feed-forward ANN is trained with Levenberg–Marquardt training algorithm. It has been shown that the developed ANN model can efficiently and accurately predict the diametral creep of pressure tube. Results show the high significance of O concentration and mechanical properties in determining diametral creep rate

  10. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    International Nuclear Information System (INIS)

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  11. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes. Additional information

    International Nuclear Information System (INIS)

    The reports from Argentina, Canada, India, Korea and Romania are presented concerning the projects carried out under the Coordinated Research Program (CRP) I3.30.10 of the International Agency for Atomic Energy - Vienna related to 'Intercomparison of Techniques for Pressure Tube Inspection and Diagnostics'

  12. The metallurgical properties of Zircaloy-2 relevant to its use in reactor pressure tubes

    International Nuclear Information System (INIS)

    The paper reviews the properties of the zirconium-base alloy Zircaloy as used in UK steam-generating heavy water reactor. It describes the major engineering requirements for this use and how the alloy can be fabricated. Some comparisons with other alloys and other metallurgical conditions are provided throughout the review. In the section on corrosion aspects the important effect of reactor irradiation in producing higher corrosion rates is emphasized by results obtained from surveillance samples taken after several years of large-scale reactor operation. In addition to oxidation effects the associated rate of hydrogen pickup is discussed and it is shown that reactor life exposure could lead to total loss of section of 0.20 mm with a total hydrogen pickup of about 100 ppm. Locally higher corrosion rates associated with nodular corrosion, and the conditions for stress corrosion are described. With regard to short-term mechanical properties, the major part of the discussion is devoted to fracture toughness aspects because of the embrittling effects of hydriding and of neutron irradiation. The experimental work shows that fast fracture could only occur under reactor conditions if a large defect was present, typically ten or more centimetres long and 80% of the wall thickness deep. Moreover, such large defects are likely to lead to leakage rather than fast fracture, satisfying 'leak-before-break' arguments. Information on fragmentation behaviour is also reported. Turning to time-dependent properties, emphasis is again put on information from long-term reactor measurements. The important aspect of irradiation-accelerated creep is reviewed with the evidence for an associated increase in strain-to-failure. It is concluded that the creep strain occurring during 25 full power years exposure to reactor conditions would not exceed 2 1/2%, a value well within the acceptable limit. (author)

  13. Fabrication and inspection of stainless-steel-clad tubes for fast reactors

    International Nuclear Information System (INIS)

    The production of cladding tubes requires a selection of the raw material, particular core taken during the cold and hot processes, special surface preparations, heat treatments, and intermediate control during the principal steps of fabrication. The inspection is made in two stages: acceptance tests at Vallourec (Eddy current and ultrasonic tests, metrology of internal and external diameter and thickness, metallography, analyses, tensile tests) and ultrasonic tests, metrology of external diameter and thickness, metallography, analyses, mechanical tests at high temperature)

  14. Derivation of Elastic Stress Concentration Factor Equations for Debris Fretting Flaws in Pressure Tubes of Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    If volumetric flaws such as bearing pad fretting flaws and debris fretting flaws are detected in the pressure tubes of pressurized heavy water reactors during in-service inspection, the initiation of fatigue cracks and delayed hydrogen cracking from the detected volumetric flaws shall be assessed by using elastic stress concentration factors in accordance with CSA N285.8-05. The CSA N285.8-05 presents only an approximate formula based on linear elastic fracture mechanics for the debris fretting flaw. In this study, an engineering formula considering the geometric characteristics of the debris fretting flaw in detail was derived using two-dimensional finite element analysis and Kinectrics, Inc.'s engineering procedure with slight modifications. Comparing the application results obtained using the derived formula with the three-dimensional finite element analysis results, it is found that the results obtained using the derived formula agree well with the results of the finite element analysis

  15. Weldability of stabilized 2 1/4Cr 1Mo Nb steel for tubes of fast reactor steam generators

    International Nuclear Information System (INIS)

    Thermorestor-W equipment was used in the study of hot and cold crackability of low-alloy chromium-molybdenum steel stabilized with niobium. The steel is designed for tubes of fast reactor steam generators. The effect of pre-heating was studied with respect to weldability. The steel is stabilized to a degree which excludes sodium-induced decarburization. Samples of three melts were used in the experiments. The steel was found not to be subject to cold crackability; however, hot crackability may be expected. It was also found that overstabilization of the steel led to an amount of eutectic sufficient for crack self-sealing. This effect, however, is not acceptable with regard to notch impact strength. In view of the possible occurrence of annealing cracks, heating rate control is recommended when the steel is heated to the annealing point. (Z.M.)

  16. KER-2 tube history

    Energy Technology Data Exchange (ETDEWEB)

    Banister, W.C.

    1963-08-16

    Zirconium process tube No. 1986 was installed in KE Reactor tube channel No. 2864 on April 16, 1959. This report describes the history and the conditions to which it was exposed during its residence in the reactor. The tube was removed on May 31, 1963.

  17. Problems of evaluation of nuclear reactor active zone tubes during pre-irradiation tests

    International Nuclear Information System (INIS)

    An analysis of standard methods of graine size estimation of basic indexes of austenitic steel and alloys of active area of atomic reactors. It is shown insolvency of standard methods of grain size estimation in the real wares. The suggested method of computer simulation of structures of pipes-shells raped for working aut of modes of heat treatment

  18. Fast reactor steam generators with sodium on the tube side. Design and operational parameters

    International Nuclear Information System (INIS)

    A comparison of design and operational characteristics as well as analysis of experience gained during the long terms operation of the Micro Module Inverse Steam Generator and Module Inverse Steam Generator at BOR 60 reactor are main aims of this technical report. 20 refs, 47 figs, 14 tabs

  19. Design Details of the CGE Vertical Heavy-Water Pressure-Tube Reactor. Some Factors Influencing the Design

    International Nuclear Information System (INIS)

    The vertical HWR basic design criteria were to simplify existing designs to improve reliability and availability and to do this with existing proven technology. It was also a requirement to maintain access to operating areas at all times and to provide adequate shutdown shielding of the reactor to permit contact maintenance. The simplified fuel channel is characterized by; (a) Single-ended fuel changing on power using only one fuelling machine. (b) End-fitting closure which stores the seal. (c) Complete factory assembly and testing of the fuel channel before on-site installation. (This is permitted because of the use of only one end fitting.) (d) Fuel channel bolted to calandria permitting easy replacement of the assembly. This feature offers the possibility of retubing the reactor when better pressure-tube materials are developed (that is, the use of thinner pressure tubes to improve burnup which could be economically attractive if the price of uranium increases) . The simplified on-power refuelling is characterized by: (a) A simplified single fuelling machine using only mechanical drives and stops. (b) The capability of reshuffling the fuel into any desired order in the string by means of an accessible fuel handling mechanism. (c) The option of direct operator control of the fuelling machine. Improvements that have been made in reducing the building volume, dry vault concept and construction procedures by using the vertical design are described. In addition, the paper describes how the station can operate at base load to take advantage of low fuelling cost or how it may be used to control system frequency. The use of reactivity control mechanisms and steam by-pass to enable the power plant to operate with a continuously varying output over a wide range is described. (author)

  20. Steam generator tube performance

    International Nuclear Information System (INIS)

    A survey of steam generator operating experience for 1986 has been carried out for 184 pressurized water and pressurized heavy-water reactors, and 1 water-cooled, graphite-moderated reactor. Tubes were plugged at 75 of the reactors (40.5%). In 1986, 3737 tubes were plugged (0.14% of those in service) and 3148 tubes were repaired by sleeving. A small number of reactors accounted for the bulk of the plugged tubes, a phenomenon consistent with previous years. For 1986, the available tubesheet sludge data for 38 reactors has been compiled into tabular form, and sludge/deposit data will be incorporated into all future surveys

  1. Steam condensation model onto horizontal finned tubes: first approximation to the containment cooling system of advanced reactors European Designs

    International Nuclear Information System (INIS)

    European designs of advanced reactors, such as EPR pr SWR 1000, have considered the use of innovative passive safety systems to preserve containment integrity even in the case of a hypothetical accident. These systems consist of several units of bundles of quasi-horizontal finned tubes. Steam released into the containment atmosphere condenses onto these structures, which are internally cooled by water under natural circulation regime. The energy absorbed by the coolant is then discharged into a pool which acts as a heat sink for at least three days. This paper presents the work carried out under the auspices of European Union within the CONGA project to simulate steam condensation onto the above mentioned quasi-horizontal finned tubes. To date calculation methodologies have been pearly reviewed and and an approximation (''Nusselt type'') has been accepted to be the most suitable for safety studies, because of its mechanistic nature and its compatibility with current safety computation tools. Two versions of this approach have been properly adapted and subsequently implemented into independent codes for their validation. An experimental database built up from the open literature allowed to point out models accuracy, showing error well within the experimental uncertainly margin. Therefore, condensate film resistance to heat transfer has been modelled satisfactorily. Nevertheless, further work remains to be done to account for the effects of noncondensable gas presence and aerosol deposition onto heat transfer surfaces. (Author) 22 refs

  2. Fuel cladding tube for nuclear reactor or method of manufacturing the same

    International Nuclear Information System (INIS)

    In a fuel cladding tube, an inner liner layer comprises, in zirconium at high purity, trace additive of one or more of tin, iron, chromium, nickel, niobium, vanadium and molybdenum, not more than 1200ppm of oxygen, and other inevitable impurities in an amount of not greater than 2000ppm in total, in which the trace additive is deposited as an intermetallic compound having an average grain size of from 0.07μm to 0.3μm inclusive. Therefore, stress corrosion cracks resistance can be improved while maintaining suppression force against abrupt oxidation at the inner surface. (T.M.)

  3. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes: Flaw detection and characterization [Phase 1

    International Nuclear Information System (INIS)

    Nuclear power plants with heavy water reactors (HWRs) comprise nine percent of today's operating nuclear units, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Programme (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the frame of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of the IAEA's project on advanced technologies for HWRs. The objective of the CRP was to inter-compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP, participants have investigated the capability of different techniques to detect and characterize flaws. During the second phase of this CRP, participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in Zirconium alloys. The intent was to identify the most effective pressure tube inspection and diagnostic methods, and to identify further development needs. The organizations that have participated in this CRP are: - The Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL); Chalk River Laboratories (CRL), Canada; - The Research Institute of Nuclear Power Operations (RINPO), China National Nuclear Corporation (CNNC), China; - Bhabha Atomic Research Centre (BARC), India; - The Korea Electric Power Research Institute (KEPRI), Republic of Korea; - The Korea Atomic Energy

  4. Environmental sciences and engineering expert to speak at Virginia Tech

    OpenAIRE

    Fay, Patrick

    2010-01-01

    Gregory Characklis of the University of North Carolina at Chapel Hill, will be speaking at Virginia Tech's Fralin Biotechnology Center Auditorium on Friday, April 23 from 9-10 a.m., to student and faculty as part of the Spring 2010 Water Seminar Series.

  5. Virginia Offshore Wind Cost Reduction Through Innovation Study (VOWCRIS) (Poster)

    Energy Technology Data Exchange (ETDEWEB)

    Maples, B.; Campbell, J.; Arora, D.

    2014-10-01

    The VOWCRIS project is an integrated systems approach to the feasibility-level design, performance, and cost-of-energy estimate for a notional 600-megawatt offshore wind project using site characteristics that apply to the Wind Energy Areas of Virginia, Maryland and North Carolina.

  6. Heat transfer burnout in tube-type fuel elements of nuclear power reactors

    International Nuclear Information System (INIS)

    The conditions are formulated under which the results of the experimental research of the boilino. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented. (F.M.)

  7. Location of grab samples from inner continental shelf of North Carolina during U.S. Geological Survey research cruises 1999-045-FA, 2001-005-FA, 2002-013-FA, 2004-003-FA (grabsamples.shp, geographic, WGS 84)

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that...

  8. Measurement of mechanical properties of a reactor operated Zr–2.5Nb pressure tube using an in situ cyclic ball indentation system

    International Nuclear Information System (INIS)

    Highlights: • Measurement of mechanical properties of pressure tube is required for its fitness assessment. • Pressure tube removal from the core consumes large amount of radiation for laboratory test. • A remotely operable In situProperty Measurement System has been designed in house. • The tool head is capable to carry out in situ ball indentation trials inside pressure tube. • The paper describes the theory and results of the trials conducted on irradiated pressure tube. - Abstract: Periodic measurement of mechanical properties of pressure tubes of Indian Pressurised Heavy Water Reactors is required for assessment of their fitness for continued operation. Removal of pressure tube from the core for preparation of specimens to test for mechanical properties in laboratories consumes large amounts of radiation and hence is to be avoided as far as possible. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing indentation test either on outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ condition. Considering this, a remotely operable hydraulic In situProperty Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed in house. The tool head of IProMS can be located inside a pressure tube at any axial location under in situ condition and the properties can be estimated from an analysis of the data on load and depth of indentation, recorded during the test. In order to qualify the system, a number of experimental trials have been conducted on spool pieces and specimens prepared from Zr–2.5Nb pressure tube having different mechanical properties. Based on the encouraging results obtained from the qualification trials, IProMS has been used inside a reactor operated

  9. Measurement of mechanical properties of a reactor operated Zr–2.5Nb pressure tube using an in situ cyclic ball indentation system

    Energy Technology Data Exchange (ETDEWEB)

    Chatterjee, S., E-mail: subrata@barc.gov.in; Panwar, Sanjay; Madhusoodanan, K.

    2015-07-15

    Highlights: • Measurement of mechanical properties of pressure tube is required for its fitness assessment. • Pressure tube removal from the core consumes large amount of radiation for laboratory test. • A remotely operable In situProperty Measurement System has been designed in house. • The tool head is capable to carry out in situ ball indentation trials inside pressure tube. • The paper describes the theory and results of the trials conducted on irradiated pressure tube. - Abstract: Periodic measurement of mechanical properties of pressure tubes of Indian Pressurised Heavy Water Reactors is required for assessment of their fitness for continued operation. Removal of pressure tube from the core for preparation of specimens to test for mechanical properties in laboratories consumes large amounts of radiation and hence is to be avoided as far as possible. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing indentation test either on outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ condition. Considering this, a remotely operable hydraulic In situProperty Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed in house. The tool head of IProMS can be located inside a pressure tube at any axial location under in situ condition and the properties can be estimated from an analysis of the data on load and depth of indentation, recorded during the test. In order to qualify the system, a number of experimental trials have been conducted on spool pieces and specimens prepared from Zr–2.5Nb pressure tube having different mechanical properties. Based on the encouraging results obtained from the qualification trials, IProMS has been used inside a reactor operated

  10. Characterization of excel alloy pressure tube material for CANDU SCW reactors

    International Nuclear Information System (INIS)

    The phase transformation temperatures, aging response, and creep rupture strength of Zr alloy Excel (Zr- 3.5%Sn- 0.8%Nb- 0.8%Mo) pressure tube material were investigated. The α → α+β and α+β → β transus temperatures were found to be in the range of 600-690 °C and 962-975 °C respectively. Precipitation hardening was observed in the microstructures water-quenched from high in the α+β or β regions followed by aging at 400-500 °C for 1 hr. The results of creep-rupture experiments at 400 °C suggest that a fully martensitic and aged microstructure has better creep properties at high stress levels (>700 MPa) and a microstructure obtained by air-cooling from high in the α+β region shows good creep properties at lower stresses (<560 MPa). (author)

  11. Effect of gadolinium nitrate concentration on the corrosion compatibility of structural materials in a proposed Indian tube type boiling reactor

    International Nuclear Information System (INIS)

    Gadolinium (Gd3+) is added with nitric acid to moderator heavy water as a neutron poison in nuclear reactors to control reactivity and pH is maintained in the range of 5 to 5.5 to prevent Gd3+ precipitation. Usually ∼15 ppm of Gd3+ is used during actuation of secondary shutdown system and is subsequently removed on ion exchange up to a residual ∼2 ppm before start-up. In the moderator system of a proposed tube type boiling water reactor of Indian origin, a higher concentration (20-400ppm) of Gd(NO3)3 was proposed to be used in the emergency safety shutdown system. With higher concentration of Gd3+, the pH can go down and affect the radiolytic yields and thus affecting the integrity of the structural materials. Considering a long life of 100 years of operation for the proposed reactor, the concentration dependence of Gd3+ on the yields of molecular products like H2 and H2O2 during radiolysis and corrosion compatibility of structural materials like (1) SS 304 LN (proposed structural material for this reactor) and (2) SS 410 (proposed to be used in the valves of the moderator system as an alternative to hard facing alloy, colmonoy) is of interest. The pH and conductivity of the system were observed to be in the range of 5.33-3.76 and 50-870 μS/cm for 20-400 ppm of Gd3+. From the electrochemical studies it was observed that the electrochemical potential increased to more positive potential with increase in Gd3+ concentrations. The yield of H2 and H2O2 was also found to increase with increase in [Gd3+] concentrations. A detailed study on corrosion of the above said alloys at varying [Gd3+] concentrations, temperature, pH and simulated irradiation conditions and its effect on microstructure will be described in the paper. (author)

  12. Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor

    International Nuclear Information System (INIS)

    Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors)

  13. The study on water ingress mass in the steam generator heat-exchange tube rupture accident of modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The steam generator heat-exchange tube rupture (SGTR) accident is an important and particular accident which will result in water ingress to the primary loop of reactor. Water ingress will result in chemical reaction of graphite fuel and structure with water, which may cause overpressure due to generation of explosive gaseous in large quantity. The study on the water ingress accident is significant for the verification of the inherent characteristics of high temperature gas-cooled reactor. The previous research shows that the amount of water ingress mass is the dominant key factor on the severity of the accident consequence. The 200 MWe high temperature gas-cooled reactor (HTR-PM), which is the first modular pebble-bed high temperature gas-cooled reactor in China designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is selected to be analyzed in this paper. The different DBA accident scenarios of double-ended break of single heat-exchange tube are simulated respectively by the thermal-hydraulic analysis code RETRAN-02. The results show the water ingress mass through the broken heat-exchange tube is related to the break location. The amount of water ingress mass is affected obviously by the capacity of the emptier system. With the balance of safety and economical efficiency, the amount of water ingress mass from the secondary side of steam generator into the primary coolant loop will be reduced by increasing properly the diameter of the draining lines. (authors)

  14. Analysis of the impact of coolant density variations in the high efficiency channel of a pressure tube super critical water reactor

    International Nuclear Information System (INIS)

    The Pressure Tube (PT) Supercritical Water Reactor (SCWR) is based on a light water coolant operating at pressures above the thermodynamic critical pressure; a separate low temperature and low pressure moderator. The coolant density changes by an order of magnitude depending on its local enthalpy in the porous ceramic insulator tube. This causes significant changes in the neutron transport characteristics, axially and radially, in the fuel channel. This work performs lattice physics calculations for a 78-element Pu-Th fuel at zero burnup and examines the effect of assumptions related to coolant density in the radial direction of a HEC, using the neutron transport code WIMS-AECL. (author)

  15. Pressure drop and heat transfer in the sodium to air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    A numerical study was performed to investigate the thermal and hydraulic characteristics and build up design model of the AHX (sodium-to-air heat exchanger) unit of a sodium-cooled fast reactor. Helical-coiled tube banks in the AHX were modeled as porous media and simulated heat and momentum transfer. Two-dimensional flow characteristic appeared at the most region of AHX annulus. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX were evaluated and compared with Zhukauskas empirical correlations. (author)

  16. Characterisation of Oxides Formed on the Internal Surface of Steam Generator Tubes in Alloy 690 Corroded in the Primary Environment of Pressurised Water Reactors

    International Nuclear Information System (INIS)

    Since the end of the 1990s, EDF R and D has been studying the phenomenon of corrosion product release from Steam Generator tubes in order to minimize the Source Term of the contamination and radiation exposure during operation and maintenance of Pressurised Water Reactors. With the BOREAL loop, release tests in primary water at 325 deg. C were performed on various Steam Generator tubes made of alloy 690. The experimental conditions of these tests (chemistry, temperature and hydraulics) were the same for all the tests but the results showed various behaviours towards release. For some tubes, the release was weak whereas for others, it was higher; the release rate of the tubes decreased more or less quickly with time. In order to explain these results, the internal surface of the tubes was characterised before and after the tests. Before the tests, various parameters were studied; the main parameters were the roughness, the impurities, the grain size and the cold work. The results demonstrated that it was not easy to quantify the influence of each parameter on release and to differentiate the tubes. A new parameter was proposed to characterise the internal extreme surface of SG tubes: the surface nano-hardness by nano-indentation measurements. The tubes were also observed and analysed by SEM, (X)TEM. Data obtained by (X)TEM revealed differences of the surface state (layer of perturbed microstructure, density of dislocations, grain size, impurities, initial oxide,...). After the tests, the oxides formed on the internal surface and the underlying material of the samples were characterised by SEM, (X)TEM and SIMS. The examinations showed various types of oxides. For some tubes, a duplex oxide scale was identified, for the others, only one oxide scale was observed. For equivalent durations of corrosion, the thickness of the enriched - chromium oxide layer can vary from 5 nm to 100 nm and the chemical composition can be different. The examinations of the underlying

  17. An Investigation on Cocombustion Behaviors of Hydrothermally Treated Municipal Solid Waste with Coal Using a Drop-Tube Reactor

    Directory of Open Access Journals (Sweden)

    Liang Lu

    2012-01-01

    Full Text Available This work aims at demonstrating the feasibility of replacing Indonesian coal (INC with hydrothermally treated municipal solid waste (MSWH in cocombustion with high ash Indian coal (IC. The combustion efficiencies and emissions (CO, NO of MSWH, INC and their blends with IC for a series of tests performed under a range of temperatures and air conditions were tested in a drop-tube reactor (DTR. The results showed the following. The combustion efficiency of IC was increased by blending both MSWH and INC and CO emission was reduced with increasing temperature. For NO emission, the blending of MSWH led to the increase of NO concentration whereas the effects of INC depended on the temperature. The combustion behaviors of IC-MSWH blend were comparable to those of the IC-INC blend indicating it is possible for MSWH to become a good substitute for INC supporting IC combustion. Moreover, the CO emission fell while the NO emission rose with increasing excess air for IC-MSWH blend at 900°C and the highest combustion efficiency was obtained at the excess air of 1.9. The existence of moisture in the cocombustion system of IC-MSWH blend could slightly improve the combustion efficiency, reduce CO, and increase NO.

  18. Thermohydraulic characteristics of serpentine tubing in the boilers of gas cooled reactors under condition of rapid and slow depressurization

    International Nuclear Information System (INIS)

    In nuclear reactors of the Magnox or advanced gas Cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accident conditions using two phase flow codes requires knowledge of the heat transfer behaviour of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear Electric. The tests were carried out on the Thermal Hydraulics Experimental Research Assembly (THERA) loop at Manchester University. Depressurization from an initial pressure of 60 bar, with fluid subcooling of 5 K, 50 K, and 100 K was controlled by discharging the test section contents through suitably chosen orifices to produce blowdown to 10% of the initial pressure over a time scale of 30 s to 3600 s. Pressures and temperatures in the serpentine were measured at average time intervals of approximately 1 s. (author)

  19. Testing of a prototype PWR design gamma thermometer for use as a local power monitor in the thimble tube of a nuclear reactor

    International Nuclear Information System (INIS)

    Out-of-pile tests were performed on a 0.185-in. OD gamma thermometer (GT) designed for insertion in existing thimble tubes of W pressurized water reactors (PWR). The experimental measurements were conducted with the GT inserted in a simulated thimble tube. The measurements were performed to examine the effect on the sensitivity, response time, and repeatability of the GT output signal when surrounded by an air- and a water-filled thimble tube with random and controlled physical contact between the two units. The results showed that the linearity of the instrument is not affected when operated in an airor water-filled thimble tube. However, the response time of the GT to a step change in power increased from 9 s to about 30 s for water and dry thimble tube conditions respectively. Results are also presented which show that the sensitivity, mV/W/cm3, of the GT with and without controlled physical contact between the inner surface of the simulated thimble tube and the outer surface of the GT did not vary significantly

  20. In-reactor Hydrogen/Deuterium ingress in Zr-2.5Nb pressure tubes and their measurement in sliver metallic samples

    International Nuclear Information System (INIS)

    Deuterium (D)/Hydrogen (H) ingress in pressure tubes (PTs) of operating PHWRs beyond solubility limit will lead to deuteride/hydride precipitation and may result in delayed hydride cracking. In-reactor H/D ingress in PTs, is periodically monitored by measuring the H/D-content in the sliver metallic samples scraped from selected PTs during their service life. This paper describes some radio analytical works on sliver samples from PTs of some of the reactors of Nuclear Power Corporation of India Limited (NPCIL). (author)

  1. Volcanoes in Virginia!

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Elizabeth Baedke [James Madison University

    2012-01-24

    The recent earthquake may have you wondering what other surprises Virginia's geology may hold. Could there be a volcanic eruption in Virginia? Probably not today, but during the Eocene, about 35-48 million years ago, a number of mysterious eruptions occurred in western Virginia. This talk investigates the possible origins of these eruptions, and what they can tell us about the crust and mantle underneath Virginia.

  2. Development of an in-service inspection technique for the intermediate heat exchanger tubes of the High-Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    An experimental study is carried out to clarify the performance of an eddy current testing probe and probe-inserting equipment for the in-service inspection of the intermediate heat exchanger tubes of the High-Temperature Engineering Test Reactor. Artificial discontinuities are made with reference to the American Society of Mechanical Engineers standards for steam generator tubes in a light water reactor. It is confirmed that the probe can detect these discontinuities as well as smaller ones, such as a 0.5-mm-diam 100% through-wall hole and a 0.5-mm-wide groove, in a base-metal tube. For the welded joints, the back-excess weld metal is a main noise contributor, and a multiple-frequency method can remove the noise. The inspection performance, however, is lower. The probe-inserting equipment can smoothly insert and extract the probe. The winding of the cable causes a scattering in the probe traveling velocity values an a measurement error regarding the probe's location in the tube

  3. Flaw detecting method for welded portion between lower end plate and stab tube of reactor pressure vessel and liquid medium-filling device used for the method

    International Nuclear Information System (INIS)

    The present invention provides a method of reliably performing an ultrasonic flaw detecting test for a welded portion between a lower end plate and a stab tube of a reactor pressure vessel. Namely, a liquid medium is filled into a space formed between the outer circumference of a housing of a driving device, and an inner surface of a driving device-insertion hole and the stab tube. Ultrasonic waves are appropriately transferred to the welded portion by means of the liquid medium. Accordingly, the ultrasonic test can reliably be performed for the welded portion between the lower end plate and the stab tube. In addition, the housing of the driving device is coated at a portion where it is situated to the outer side of the main body of the pressure vessel. The liquid medium is continuously supplied from a medium supply device into the inside of the main body of the filling device. The liquid medium is filled into the space formed by the outer circumference of the housing of the driving device, and the inner surface of the insertion hole of the driving device and the stab tube. Accordingly, ultrasonic test can reliably performed for the welded portion between the lower end plate and the stab tube. (I.S.)

  4. Combination of helical ferritic-steel inserts and flux-tube-expansion divertor for the heat control in tokamak DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takizuka, T., E-mail: takizuka.tomonori@gmail.com [Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita 565-0871 (Japan); Tokunaga, S.; Hoshino, K. [Japan Atomic Energy Agency, 2-166, Omotedate, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Shimizu, K. [Japan Atomic Energy Agency, 801-1, Mukoyama, Naka 311-0193 (Japan); Asakura, N. [Japan Atomic Energy Agency, 2-166, Omotedate, Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2015-08-15

    Edge localized modes (ELMs) in the H-mode operation of tokamak reactors may be suppressed/mitigated by the resonant magnetic perturbation (RMP), but RMP coils are considered incompatible with DEMO reactors under the strong neutron flux. We propose an innovative concept of the RMP without installing coils but inserting ferritic steels of the helical configuration. Helically perturbed field is naturally formed in the axisymmetric toroidal field through the helical ferritic steel inserts (FSIs). When ELMs are avoided, large stationary heat load on divertor plates can be reduced by adopting a flux-tube-expansion (FTE) divertor like an X divertor. Separatrix shape and divertor-plate inclination are similar to those of a simple long-leg divertor configuration. Combination of the helical FSIs and the FTE divertor is a suitable method for the heat control to avoid transient ELM heat pulse and to reduce stationary divertor heat load in a tokamak DEMO reactor.

  5. Theoretical and experimental modeling of processes accompanying single fuel channel (pressure tube) rupture for RBMK Reactor. Part A

    International Nuclear Information System (INIS)

    The analysis of the MPTR (Multiple Pressure Tube Rupture) problem requires a series of theoretical and experimental studies of separate physical processes in the RBMK reactor, development of mathematical models and their physical equivalents. The experimental rigs concerned with MPTR problem were designed and constructed at Electrogorsk Research and Engineering Center, Russia. An investigation of the circumstances and mechanism of a rupture in a single channel at different conditions and scenarios is one of the main stages of the aforesaid analysis. Theoretical models of the single channel rupture under thermal and mechanical loading was developed including a channel constrained graphite block. Computer program based on this models enable to describe thermomechanical deformation process of the single channel and to predict rupture moment. Theoretical investigations supplements with experimental modelling single channel rupture by means of series experimental examinations at TKR-F (Model of an Accidental Channel) test rig. One represents a model of the single ruptured fuel channel in a surrounding graphite column. Experimental examinations enable to develop and verify theoretical models, conditions and mechanism of a rupture in a single channel. The flow process of steam or steam-water mixture through narrow graphite gaps is another important process should be modelled in frame of the MPTR problem analysis. The TKR-F (graphite) facility represents two test sections: for high speeds of medium flow near the ruptured channel and for low speeds far from the rupture. In the first case dynamical interaction escaping steam from the ruptured channel with (moving) graphite columns is modelled. The second case is attended to studying of hydraulical resistances and heat transfer of the steady-state steam-water flow within graphite gaps with different cross-section. Theoretical and experimental modelling consolidation sets out technique of the authentic analysis of the

  6. The application of an innovative continuous multiple tube reactor as a strategy to control the specific organic loading rate for biohydrogen production by dark fermentation.

    Science.gov (United States)

    Gomes, Simone D; Fuess, Lucas T; Penteado, Eduardo D; Lucas, Shaiane D M; Gotardo, Jackeline T; Zaiat, Marcelo

    2015-12-01

    Biohydrogen production in fixed-bed reactors often leads to unstable and decreasing patterns because the excessive accumulation of biomass in the bed negatively affects the specific organic loading rate (SOLR) applied to the reactor. In this context, an innovative reactor configuration, i.e., the continuous multiple tube reactor (CMTR), was assessed in an attempt to better control the SOLR for biohydrogen production. The CMTR provides a continuous discharge of biomass, preventing the accumulation of solids in the long-term. Sucrose was used as the carbon source and mesophilic temperature conditions (25°C) were applied in three continuous assays. The reactor showed better performance when support material was placed in the outlet chamber to enhance biomass retention within the reactor. Although the SOLR could not be effectively controlled, reaching values usually higher than 10gsucroseg(-1)VSSd(-1), the volumetric hydrogen production and molar hydrogen production rates peaked, respectively, at 1470mLH2L(-1)d(-1) and 45mmolH2d(-1), indicating that the CMTR was a suitable configuration for biohydrogen production. PMID:26340028

  7. A repair process for an heterogenous welded joint between a nuclear reactor component tube and a pipe

    International Nuclear Information System (INIS)

    The repairing process involves cutting a tubular section of the tube and the pipe, which includes the welded joint, and preparing an austenitic stainless steel tubular section for substitution; the section is then narrow-joint welded with the low-alloy steel tube, and finally welded to the austenitic stainless steel pipe. Application to repairing a welded joint between a PWR pressurizer tube and the expansion pipe of the pressurizer. (authors). 7 refs., 3 figs

  8. Use of a packed-bed airlift reactor with net draft tube to study kinetics of naphthalene degradation by Ralstonia eutropha.

    Science.gov (United States)

    Jalilnejad, Elham; Vahabzadeh, Farzaneh

    2014-03-01

    Biodegradation of naphthalene by Ralstonia eutropha (also known as Cupriavidus necator) in a packed-bed airlift reactor with net draft tube (PBALR-nd) was studied; the Kissiris pieces were the packing material. The reactor hydrodynamics has been characterized under abiotic conditions and the dependencies of the superficial gas velocity (U G) on the gas holdup (εG), liquid mixing time, and mass transfer coefficient were determined. The improving role of the net draft tube in this small column reactor (height 42 cm, ID 5 cm) was confirmed. The flow regime was described using the εG α U G (n) expression, and bubbly flow was observed in PBALR-nd at U G < 2.83 cm/s. In the second step of the present work, the kinetics of biodegradation was modeled using the Haldane and Aiba equations. The fitting of the experimental results to the models were done according to the nonlinear least square regression technique. The biokinetic constants (q m, K s, and K i) were estimated and q m as the specific biodegradation rate was equaled to 0.415 and 0.24 mgnaph./mgcell h for the Haldane and Aiba equations, respectively. The goodness of fit reported as R (2) and root-mean-square error (RMSE) showed the adequate fitness of the Haldane and Aiba models in predicting naphthalene biodegradation kinetics. On the basis of the HPLC results, a hypothetical pathway for the biodegradation was presented. PMID:24338109

  9. Biaxial creep deformation behavior of Fe–14Cr–15Ni–Ti modified austenitic stainless steel fuel cladding tube for sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Highlights: • Significant amounts of creep strain is observed in the axial and hoop directions. • Hoop strain is much higher than the axial strain. • Steady state hoop rate is lower than steady state axial rate. • Steady state hoop rate is comparable with creep rate evaluated from uniaxial tests. • Alloy D9 exhibits anisotropy in creep deformation. - Abstract: Twenty percent cold worked Fe–14Cr–15Ni–Ti modified austenitic stainless steel is used as the cladding tube material for the fuel pins of the Prototype Fast Breeder Reactor in India. Biaxial creep properties of the tubes have been studied at 973 K by carrying out creep tests by internally pressurizing the tubes. Hoop and axial components of creep strain were measured and found to be significantly different. For a given gas pressure, steady state hoop rate was higher than the axial rate. Steady state hoop and axial creep rates followed Norton's power law with the same stress exponent n = 7. Steady state hoop rates determined from biaxial creep tests agreed with the steady state creep rates determined from uniaxial creep tests. For a thin walled closed tube under internal pressure, significant axial deformation along with hoop deformation is indicative of anisotropic deformation of the material

  10. Case study of the failure of supercritical water oxidation reactor tubing during the treatment of 2,4 DNP with ammonium sulphate

    International Nuclear Information System (INIS)

    During the process of Supercritical Water Oxidation (SCWO) organic chemical streams are oxidized at high temperature and pressure, typically in excess of 647 K and 22.1 MPa. Due to high operating temperatures and pressures severely corrosive environments often ensue and eventually lead to SCWO reactor tube failures. This case study looked at one such failure of Alloy 625 (61Ni-21.5Cr-9Mo) tubing which occurred at the UBC/NORAM SCWO pilot plant while treating a feed of waste water containing 2.4 wt% 2,4 dinitrophenolate, 2% ammonium sulphate and 6% excess ammonia. Although the feed pH was approximately 9 and therefore not expected to be corrosive, in fact the tube failed when exposed to this feed (with oxygen) for a period of about 1 hour at 650-655 K. Through the examination of the ensuing thermodynamic system as well as SEM and Optical Microscope analysis of the ruptured portions of tubing, it was found that the addition of ammonium sulphate to the treated media caused rapid failure due to de-alloying. Findings show that the high sub-critical temperature and high density of the feed water at failure points, as well as the corrosion morphology are consistent with attack by ammonia. The formation of a stable soluble nickel-ammine phase is suspected. (author)

  11. Biaxial creep deformation behavior of Fe–14Cr–15Ni–Ti modified austenitic stainless steel fuel cladding tube for sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mathew, M.D., E-mail: mathew@igcar.gov.in [Mechanical Metallurgy Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Ravi, S.; Vijayanand, V.D.; Latha, S. [Mechanical Metallurgy Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Dasgupta, Arup [Physical Metallurgy Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Laha, K. [Mechanical Metallurgy Division, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

    2014-08-15

    Highlights: • Significant amounts of creep strain is observed in the axial and hoop directions. • Hoop strain is much higher than the axial strain. • Steady state hoop rate is lower than steady state axial rate. • Steady state hoop rate is comparable with creep rate evaluated from uniaxial tests. • Alloy D9 exhibits anisotropy in creep deformation. - Abstract: Twenty percent cold worked Fe–14Cr–15Ni–Ti modified austenitic stainless steel is used as the cladding tube material for the fuel pins of the Prototype Fast Breeder Reactor in India. Biaxial creep properties of the tubes have been studied at 973 K by carrying out creep tests by internally pressurizing the tubes. Hoop and axial components of creep strain were measured and found to be significantly different. For a given gas pressure, steady state hoop rate was higher than the axial rate. Steady state hoop and axial creep rates followed Norton's power law with the same stress exponent n = 7. Steady state hoop rates determined from biaxial creep tests agreed with the steady state creep rates determined from uniaxial creep tests. For a thin walled closed tube under internal pressure, significant axial deformation along with hoop deformation is indicative of anisotropic deformation of the material.

  12. Tests for development of estimation technology of reactor core deformation. Report No.1: fundamental mechanical properties of wrapper tube (test report)

    Energy Technology Data Exchange (ETDEWEB)

    Nishiura, Takeo; Shimazaki, Yuji; Horikiri, Morito [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-10-01

    Mechanical properties such as local contact compression stiffness, bending stiffness, deformation properties, material properties, and friction properties of a wrapper tube structure were clarified experimentally, which can be used as the basic data for development of estimation technology of reactor core deformation. Contents of the Tests data as follows: (1) Effects of load supporting boundary conditions, whether or not a contact-proof pad is attached, and length of duct, on cross section deformation of wrapper tube were made clear as the local contact compression stiffness characteristics. (2) Bending stiffness does not depend on the difference of load supporting boundary conditions. The property of cross section deformation under bending load was obtained. (3) The deformation modes and the strain distributions were obtained by the deformation tests of wrapper tube. (4) The stress-strain diagrams including plastic range under various strain variation rates were obtained by the material tests at room temperature. (5) The static and the dynamic friction coefficients by various contact angles and the contact loads between contact-proof pads of two wrapper tubes were obtained by friction property tests. (author)

  13. A repair process for an heterogenous welded joint between a nuclear reactor component tube and a pipe

    International Nuclear Information System (INIS)

    The repairing process involves cutting a tubular section of the tube (made of low alloy steel) and the pipe (made of austenitic stainless steel), which includes the welded joint, and preparing an heterogenous tubular section for substitution (a first section, made of ferritic steel, is butt welded to a second section, made of austenitic stainless steel); the tubular section is then narrow-joint welded with the low-alloy steel tube, and finally welded to the austenitic stainless steel pipe. Application to repairing a welded joint between a pressurizer tube and an expansion pipe connected to the primary circuit. (author). 5 refs., 4 figs

  14. Methodical and experimental features of operative periodic control of mechanic properties in metal tubes of WWER-1000 reactors

    International Nuclear Information System (INIS)

    The results of the methodical development and the results of the base material of the weld metal mechanical properties control for the RU WWER-1000 tubing after 100 thousand hours of service are generalized. The problem of the tubing service life extension for the operating Power Station blocks of WWER-1000 of Russia and Ukraine is considered on the base of the control of mechanical properties that depend on ageing

  15. Biodegradation of 2,4,6-trichlorophenol in a packed-bed biofilm reactor equipped with an internal net draft tube riser for aeration and liquid circulation

    International Nuclear Information System (INIS)

    For the aerobic biodegradation of the fungicide and defoliant 2,4,6-trichlorophenol (2,4,6-TCP), a bench-scale packed-bed bioreactor equipped with a net draft tube riser for liquid circulation and oxygenation (PB-ALR) was constructed. To obtain a high packed-bed volume relative to the whole bioreactor volume, a high AD/AR ratio was used. Reactor's downcomer was packed with a porous support of volcanic stone fragments. PB-ALR hydrodynamics and oxygen mass transfer behavior was evaluated and compared to the observed behavior of the unpacked reactor operating as an internal airlift reactor (ALR). Overall gas holdup values εG, and zonal oxygen mass transfer coefficients determined at various airflow rates in the PB-ALR, were higher than those obtained with the ALR. When comparing mixing time values obtained in both cases, a slight increment in mixing time was observed when reactor was operated as a PB-ALR. By using a mixed microbial community, the biofilm reactor was used to evaluate the aerobic biodegradation of 2,4,6-TCP. Three bacterial strains identified as Burkholderia sp., Burkholderia kururiensis and Stenotrophomonas sp. constituted the microbial consortium able to cometabolically degrade the 2,4,6-TCP, using phenol as primary substrate. This consortium removed 100% of phenol and near 99% of 2,4,6-TCP. Mineralization and dehalogenation of 2,4,6-TCP was evidenced by high COD removal efficiencies (∼95%), and by the stoichiometric release of chloride ions from the halogenated compound (∼80%). Finally, it was observed that the microbial consortium was also capable to metabolize 2,4,6-TCP without phenol as primary substrate, with high removal efficiencies (near 100% for 2,4,6-TCP, 92% for COD and 88% for chloride ions)

  16. Biodegradation of 2,4,6-trichlorophenol in a packed-bed biofilm reactor equipped with an internal net draft tube riser for aeration and liquid circulation

    Energy Technology Data Exchange (ETDEWEB)

    Gomez-De Jesus, A.; Romano-Baez, F.J.; Leyva-Amezcua, L.; Juarez-Ramirez, C.; Ruiz-Ordaz, N. [Departamento de Ingenieria Bioquimica, Escuela Nacional de Ciencias Biologicas, IPN. Prol. Carpio y Plan de Ayala, Colonia Santo Tomas, s/n. CP 11340, Mexico, D.F. (Mexico); Galindez-Mayer, J. [Departamento de Ingenieria Bioquimica, Escuela Nacional de Ciencias Biologicas, IPN. Prol. Carpio y Plan de Ayala, Colonia Santo Tomas, s/n. CP 11340, Mexico, D.F. (Mexico)], E-mail: cmayer@encb.ipn.mx

    2009-01-30

    For the aerobic biodegradation of the fungicide and defoliant 2,4,6-trichlorophenol (2,4,6-TCP), a bench-scale packed-bed bioreactor equipped with a net draft tube riser for liquid circulation and oxygenation (PB-ALR) was constructed. To obtain a high packed-bed volume relative to the whole bioreactor volume, a high A{sub D}/A{sub R} ratio was used. Reactor's downcomer was packed with a porous support of volcanic stone fragments. PB-ALR hydrodynamics and oxygen mass transfer behavior was evaluated and compared to the observed behavior of the unpacked reactor operating as an internal airlift reactor (ALR). Overall gas holdup values {epsilon}{sub G}, and zonal oxygen mass transfer coefficients determined at various airflow rates in the PB-ALR, were higher than those obtained with the ALR. When comparing mixing time values obtained in both cases, a slight increment in mixing time was observed when reactor was operated as a PB-ALR. By using a mixed microbial community, the biofilm reactor was used to evaluate the aerobic biodegradation of 2,4,6-TCP. Three bacterial strains identified as Burkholderia sp., Burkholderia kururiensis and Stenotrophomonas sp. constituted the microbial consortium able to cometabolically degrade the 2,4,6-TCP, using phenol as primary substrate. This consortium removed 100% of phenol and near 99% of 2,4,6-TCP. Mineralization and dehalogenation of 2,4,6-TCP was evidenced by high COD removal efficiencies ({approx}95%), and by the stoichiometric release of chloride ions from the halogenated compound ({approx}80%). Finally, it was observed that the microbial consortium was also capable to metabolize 2,4,6-TCP without phenol as primary substrate, with high removal efficiencies (near 100% for 2,4,6-TCP, 92% for COD and 88% for chloride ions)

  17. Remotely controlled device for tightening, the nuts on locating pins for guide tubes in pressurized water reactors

    International Nuclear Information System (INIS)

    The device has a support having a horizontal guide radial to the guide tube with a trolley moving on the guide and mounted on it a tool carrier. The tightening tool it self consists of a motor and an assembly of reducing gears mounted on the tool carrier. The final gear wheel in the assembly turns about a vertical axis and has a ferrule on its face for tightening the nut of the guide tube locating pin. The force of reaction on the tool carrier may be measured thus allowing the torque applied by the tool to be regulated

  18. Simulation of steam condensation in the presence of noncondensable gases in horizontal condenser tubes using RELAP5 for advanced nuclear reactors

    International Nuclear Information System (INIS)

    Horizontal heat exchangers are used in advanced light water nuclear reactors in their passive cooling systems, such as residual heat removal (RHRS) and passive containment cooling system (PCCS). Condensation studies of steam and noncondensable gases mixtures in these heat exchangers are very important due to the phenomena multidimensional nature and the condensate stratification effects. This work presents a comparison between simulation results and experimental data in steady state conditions for some inlet pressure, steam and noncondensable gases (air) inlet mass fractions. The test section is three meters long and consists of two concentric tubes containing pressure, temperature and flow rate sensors. The internal tube, called condenser, contains steam-air mixture flow and external tube is a counter current cooler with water flow rate at low temperature. This test section was modeled and simulations were performed with RELAP5 code. Experimental tests were carried out for 200 to 400 kPa inlet pressure and 5, 10, 15 and 20% of inlet air mass fractions. Comparisons between experimental data and simulation results are presented for 200 and 400 kPa pressure conditions and showed good agreement. However, for 400 kPa inlet steam pressure and inlet air mass fractions above 5%, the simulated temperatures are lower than the experimental data at the final third from the inlet condenser tube, indicating a code overestimation of heat transfer coefficient. New correlations for heat transfer coefficient in these steam-air conditions must be theoretical and experimentally studied and implemented in RELAP5 code for better representing the condensation phenomena. (author)

  19. Census Snapshot: Virginia

    OpenAIRE

    Romero, Adam P; Rosky, Clifford J; Badgett, M. V. Lee; Gates, Gary J.

    2008-01-01

    Using data from the U.S. Census Bureau, this report provides demographic and economic information about same-sex couples and same-sex couples raising children in Virginia. We compare same-sex “unmarried partners,” which the Census Bureau defines as an unmarried couple who “shares living quarters and has a close personal relationship,” to different-sex married couples in Virginia. In many ways, the almost 20,000 same-sex couples living in Virginia are similar to married couples. Accor...

  20. Census Snapshot: West Virginia

    OpenAIRE

    Romero, Adam P; Rosky, Clifford J; Badgett, M. V. Lee; Gates, Gary J.

    2008-01-01

    Using data from the U.S. Census Bureau, this report provides demographic and economic information about same-sex couples and same-sex couples raising children in West Virginia. We compare same-sex “unmarried partners,” which the Census Bureau defines as an unmarried couple who “shares living quarters and has a close personal relationship,” to different-sex married couples in West Virginia. In many ways, the almost 3,500 same-sex couples living in West Virginia are similar to married ...

  1. Steam generator tube failures

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, P.E.; Shah, V.N.; Ward, L.W.; Ellison, P.G.

    1996-04-01

    A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service.

  2. Steam generator tube failures

    International Nuclear Information System (INIS)

    A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service

  3. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2006-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  4. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2005-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  5. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2012-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  6. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    1999-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  7. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2001-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  8. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2000-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  9. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    1998-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  10. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2003-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  11. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2007-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  12. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2010-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  13. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2009-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  14. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2011-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  15. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2008-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  16. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2004-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  17. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2013-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  18. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2002-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  19. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    2014-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia.

  20. Straightening tubes

    International Nuclear Information System (INIS)

    Hexagonal wrapper tubes, especially for nuclear reactor core sub-assemblies, may suffer from unacceptable bow as a result of welding wear pads to the wrapper and heat treatment. Straightening of the bow is effected by a method wherein at each of a series of axially spaced locations the faces or vertices of the tube are measured relative to a reference to determine the direction of bow at the locations. From these measurements, the appropriate axial locations for the application of corrective loading can be determined, whereby by application of the loading at a selected face or vertex for such measurements the bow is reduced. Such loading, by an actuator, can be repeated at the locations until the bow is reduced to within tolerances. (author)

  1. Hydrology of the Dismal Swamp Virginia and North Carolina 1974

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — The purpose of this report is to summarize and interpret information on the hydrology of the Dismal Swamp area and the related geology and to suggest aspects of the...

  2. Selection of instruments used for vibration measurement of fuel bundles in a pressure tube under CANDU reactor operating conditions

    International Nuclear Information System (INIS)

    Vibration characteristics of CANDU fuel bundle and fuel elements is a key parameter considered in the design of a fuel bundle. Out-reactor frequency and temperature sweep tests, under reactor operating conditions, are performed to verify vibration characteristics of CANDU fuel bundles. Several options have been considered in the selection of vibration instrumentation to perform out-reactor frequency and temperature sweep tests. This paper compares the benefits and disadvantages of various vibration instruments and summarizes the rationale behind the selection of instruments used for vibration measurements over a range of temperature and pressure pulsation frequencies. The conclusions are presented from the bench tests performed, which confirm the use of the selected instruments. (author)

  3. Analysis of multiple-tube ruptures in both steam generators for the Three Mile Island-1 pressurized water reactor

    International Nuclear Information System (INIS)

    The operator guidelines were followed for both transients described. Both transients resulted in SG overfill and the tube-rupture flow did not terminate in either transient. The following statements can be deducted from the results of the calculations: the tube-rupture flow could not be stopped for either case during 2600 s (43 min) of transient time; each accident scenario resulted in SG overfill; both SGs overfilled by 1600 s (27 min) and 1800 s (30 min) for Cases 1 and 2, respectively; conditions for isolation of the SGs were not reached; and core subcooling was not lost in either case but the upper head was voided in Case 2. Comparison of the cooldown rates in the two cases after 1200 s (20 min) shows that these rates are equal (i.e., restart of the RCPs did not change the primary-system cooldown rate). However, in Case 2, a steam bubble was formed in the upper head, which did not disappear during the simulated time. One of the immediate actions in the guidelines was to fill both SGs to 95% level. This step was almost unnecessary because the tube-rupture flow was large enough that the 15.4 - mK/s (100 - 0F h) cooldown-rate limit was exceeded and AFW could not be injected. Also, the guidelines did not address the SG overfill issue

  4. Fuel composition optimization in a 78-element fuel bundle for use in a pressure tube type supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    A 78-element fuel bundle containing a plutonium-thorium fuel mixture has been proposed for a Generation IV pressure tube type supercritical water-cooled reactor. In this work, using a lattice cell model created with the code DRAGON,the lattice pitch, fuel composition (fraction of PuO2 in ThO2) and radial enrichment profile of the 78-element bundle is optimized using a merit function and a metaheuristic search algorithm.The merit function is designed such that the optimal fuel maximizes fuel utilization while minimizing peak element ratings and coolant void reactivity. A radial enrichment profile of 10 wt%, 11 wt% and 20 wt% PuO2 (inner to outer ring) with a lattice pitch of 25.0 cm was found to provide the optimal merit score based on the aforementioned criteria. (author)

  5. Thermo-hydraulic characteristics of serpentine tubing in the boilers of gas cooled reactors under condition of rapid and slow depressurization

    International Nuclear Information System (INIS)

    In nuclear reactors of the magnox or advanced gas cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accidents using two phase flow codes requires knowledge of the heat transfer behaviour of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear electric . The tests were carried out on the thermal hydraulics experimental research assembly (THERA) loop at manchester university. Depressurization from an initial pressure of 60 bar, with fluid subcooling of 5 k, 50 k, and 100 k was controlled by discharging the test section contents through suitably chosen orifices to produce blowdown to 10% of the initial pressure over a time scale of 30 s to 3600 s. pressures and temperatures in the serpentine were measured at average time intervals of approximately 1 s

  6. 75 FR 39285 - Virginia Electric and Power Company: North Anna Power Station, Unit No. 1 Environmental...

    Science.gov (United States)

    2010-07-08

    ... COMMISSION Virginia Electric and Power Company: North Anna Power Station, Unit No. 1 Environmental Assessment..., Section III.O, ``Oil collection system for reactor coolant pump,'' for Facility Operating License No. NPF... Power Station, Unit 1 (NAPS Unit 1), located in Louisa County, Virginia. Therefore, as required by...

  7. Taking into account steam generator tube failure accidents in the design of safety systems for the N4 reactor (PWR 1400)

    International Nuclear Information System (INIS)

    Owing to the frequency with which they occur on PWRs, steam generator tube failure accidents have been the subject of a comprehensive review in France. Many modifications have been incorporated into the new PWR 1400 reactors so as to take this kind of accident more fully into account. The aim of these modifications is to control as well as possible the discharge of steam to the atmosphere, on the one hand, by making the atmospheric dump system an engineered safety feature, and on the other hand, by keeping to a minimum the risk of stressing the steam generator blow-off valves. Other modifications have been made to the turbine trip and to the startup of the emergency feedwater system, with a view to limiting the risk of an uncontrolled discharge into the atmosphere. The approach adopted represents an example of the integration of feedback into the design of safety systems and makes it possible to improve the level of safety in relation not only to steam generator tube failure accidents, but to all accidents which would necessitate rapid cooling by the secondary system. (author). 1 fig., 2 tabs

  8. Experimental qualification of the calculation methods of fluid-structure interactions in the tube circuits of nuclear reactors

    International Nuclear Information System (INIS)

    This work concerns the theoretical and experimental study of the propagation and damping of pressure waves (pressure losses), generated by an unsteady flow (sinusoidally oscillating flow) passing through singularities (constructions having orifices with square edges) placed in ducts. After a systematic review of state-of-the-art of losses in unsteady flows, a preliminary study of the behavior of a variable-pressure source, generated by the vibrating plate, in a fluid was completed. In order to perform the experimental measurements we have realized and exploited a specifically designed apparatus, consisting of a piston, driven by a vibrator, which interacts with a fluid in a horizontal tube. In the test section of the flow circuit, the singularity was introduced by interposing perforated plates, of variable geometry, in the tube. The pressure profiles are measured along the tube. A mathematical model, based on the hydrodynamic equations has been developed to interpret the experimental observations. This study has shown that the pressure drop between the two sides of the singularity can be accounted for. The period parameter was varied from about 0.2 to 23, the Reynolds number from 5 x 102 to 1.2 x 105, and the section porosity (orifice/pipe area ratio) from 0.05 to 0.5. We have not found any evident correlation between the pressure loss coefficient and Reynolds number. For high values of the period parameter, the pressure loss coefficient approaches its predicted value by the steady flow; in the low values, a wide scatter of measured pressure loss coefficient was found to be consistently greater than those resulting in the steady flow case

  9. Census Snapshot: North Carolina

    OpenAIRE

    Romero, Adam P; Rosky, Clifford J; Badgett, M. V. Lee; Gates, Gary J.

    2008-01-01

    Using data from the U.S. Census Bureau, this report provides demographic and economic information about same-sex couples and same-sex couples raising children in North Carolina. We compare same-sex “unmarried partners,” which the Census Bureau defines as an unmarried couple who “shares living quarters and has a close personal relationship,” to different-sex married couples in North Carolina. In many ways, the nearly 20,000 same-sex couples living in North Carolina are similar to marr...

  10. Census Snapshot: South Carolina

    OpenAIRE

    Romero, Adam P; Rosky, Clifford J; Badgett, M. V. Lee; Gates, Gary J.

    2008-01-01

    Using data from the U.S. Census Bureau, this report provides demographic and economic information about same-sex couples and same-sex couples raising children in South Carolina. We compare same-sex “unmarried partners,” which the Census Bureau defines as an unmarried couple who “shares living quarters and has a close personal relationship,” to different-sex married couples in South Carolina. In many ways, the more than 10,500 same-sex couples living in South Carolina are similar to married c...

  11. Automatic thickness measuring system of zirconium and zircaloy-2 layers of zirconium liner cladding tubes for boiling water reactor

    International Nuclear Information System (INIS)

    An automatic measuring system using ultrasonic method and electromagnetic method has been developed to measure the thickness of zirconium and zircaloy-2 layers. The sophisticated mechanism and the unique signal processing for suppression of several types of error enable high accurate measurement. The standard deviation of the liner thickness measurement is 2.2 μm and that of mother layer measurement is 3.0 μm. This system is very useful to assure the thickness of each layer and to produce high quality zirconium liner cladding tubes. (author)

  12. Implementation of Wolsong Pump Model, Pressure Tube Deformation Model and Off-take Model into MARS Code for Regulatory Auditing of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Y. J.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2008-05-15

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes.

  13. Implementation of Wolsong Pump Model, Pressure Tube Deformation Model and Off-take Model into MARS Code for Regulatory Auditing of CANDU Reactors

    International Nuclear Information System (INIS)

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes

  14. COD, 2,4,6-trichlorophenol (TCP) and toxicity removal from synthetic wastewater in a rotating perforated-tubes biofilm reactor

    International Nuclear Information System (INIS)

    Synthetic wastewater containing different concentrations of 2,4,6-trichlorophenol (TCP) was biologically treated using a novel rotating perforated-tubes biofilm reactor (RTBR) for chemical oxygen demand (COD), TCP and toxicity removal. Performance of the reactor was investigated as function of major operating variables such as the feed TCP and COD concentrations and A/Q (biofilm surface area/feed flow rate) ratio. A Box-Behnken statistical experiment design method was used by considering the feed TCP (0-400 mg L-1), COD (1000-4000 mg L-1) and A/Q ratio (23-163 m2 d m-3) as the independent variables while percent TCP, COD, and toxicity removals were the objective functions. The results were correlated with the quadratic model since this was found to be the most suitable one. Response function coefficients were determined by correlating the experimental data with the response function. Percent TCP, COD and toxicity removals estimated from the response functions were in good agreement with the experimental results. TCP, COD and toxicity removals increased with increasing A/Q ratio and decreasing feed TCP concentrations. Percent toxicity removals were always lower than TCP removals indicating presence or formation of some toxic by products from TCP biodegradation. For the feed TCP of 400 mg L-1, the optimum conditions resulting in maximum COD (99%), TCP (100%) and toxicity (93%) removals were A/Q ratio of nearly 165 m2 d m-3 and feed COD of 2985 mg L-1

  15. Virginia Beginning Farmer & Rancher Coalition Program: Virginia Beginning Farmer Profiles

    OpenAIRE

    Niewolny, Kim; Whitter-Cummings, Althea

    2013-01-01

    The purpose of the Virginia Beginning Farmer Profiles is to help beginning farmers and ranchers in Virginia gain knowledge about farm start-up and planning from first-hand, personal experiences shared by other beginning farmers. This educational resource is component of the Virginia Beginning Farmer and Rancher Coalition Program's "Whole Farm Planning" curriculum.

  16. Virginia Atlantic Coast Recreational Use

    Data.gov (United States)

    Virginia Department of Environmental Quality — As a member of the Mid-Atlantic Regional Council on the Ocean (MARCO), Virginia, through its Coastal Zone Management (CZM) Program, collected information on how the...

  17. Progress and challenges in low-level waste disposal in North Carolina

    International Nuclear Information System (INIS)

    Eight states in the Southeastern US responded to the Low-Level Radioactive Waste Policy Act of 1980 by forming the Southeast Compact. Included are Alabama, Florida, Georgia, Mississippi, North Carolina, South Carolina, Tennessee, and Virginia. The Compact commissioners in 1986 selected North Carolina as the second host state for a disposal facility using a procedure based on state features and capabilities. South Carolina is the first host state. The North Carolina General Assembly in 1987 created the North Carolina Low-Level Radioactive Waste management Authority (the Authority), a 15-member citizen board. The Authority was given responsibilities, powers, and a target schedule to site, operate, close, and maintain the disposal facility. The North Carolina radiation Protection Commission issued rules based on NRC's 10CFR61 and unique North Carolina law. Shallow land burial is banned, and engineered barriers are required. The project has been delayed by several factors and events: lengthy site operator contract negotiations, the insertion of the precharacterization step, the extended review of the site characterization plans, and litigation by Richmond and Chatham counties. Complaints include allegations that an environmental impact statement (EIS) is needed before site characterization can begin and that the Authority has failed to comply fully with the law during the site-screening process. As a result of delays, the facility is not likely to be open for waste disposal until 1995. With Barnwell scheduled to close December 31, 1992, generators must store wastes for ∼2 years or make other arrangements

  18. Manipulator for inspection or repair of heat exchanger tubes, in particular in steam generators for nuclear reactors

    International Nuclear Information System (INIS)

    The manipulator used to inspect or repair pipes in the steam generator chamber of a PWR can be introduced and removed through a penetration nozzle which can be sealed tightly by means of a blind flange. The front end of the manipulator carries a swivel arm which can be operated remotely to be moved in a plane parallel to the tube plate. The end of the swivel arm carries a holder for a mouthpiece which can be extended and retracted. This carrier can also be operated remotely so as to be aligned to the pipe orifices in a direction normal to the swivel plane of the swivel arm. The manipulator is supported in antifriction bearings in the penetration nozzle so as to be movable longitudinally. (DG)

  19. 75 FR 3942 - Carolina Power & Light Company Shearon Harris Nuclear Power Plant, Unit 1 Environmental...

    Science.gov (United States)

    2010-01-25

    ... impact (Part 73, Power Reactor Security Requirements, 74 FR 13926 through 13967, dated March 27, 2009... Carolina Power & Light Company (the licensee), now doing business as Progress Energy Carolinas, Inc. (PEC... promulgating its revisions to 10 CFR Part 73 as discussed in a Federal Register (FR) notice dated March...

  20. Design and preliminary results of an NMR tube reactor to study the oxidative degradation of fatty acid methyl ester

    International Nuclear Information System (INIS)

    Biodiesel is the fatty acid alkyl esters produced by the transesterification of vegetable, animal or microbial lipids. After ethanol, it accounts for the largest proportion of global biofuel production. Yet, due to the level of polyunsaturation, biodiesel is also oxidatively unstable. When biodiesel oxidises the viscosity increases, which leads to reduced fuel performance and in extreme cases can lead to engine failure. To aid in understanding the process of this degradation a specialist NMR tube rig was designed to assess the oxidation of biodiesel. The NMR tube rig allowed the in situ1H NMR measurement of the sample while air was bubbled through at fixed intervals. The methyl esters of linolenic acid (18:3), linoleic acid (18:2) and oleic acid (18:1) were oxidised at 110 °C over a 24 h period. The decomposition of biodiesel is complex, and there is more than one mechanism involved in the degradation. Using this rig the onset of oxidation for 18:3 and 18:2 was found to be almost instantaneous. The rate of oxidation was found to be slightly less for 18:2 than 18:3 while the maximum rate was observed for 18:3 from the beginning of the oxidation, this was only observed after 280 min for 18:2. The oxidation of 18:1 started at approximately 500 min and, slowly degraded during the remaining reaction time. The formation of a number of secondary oxidation products such as aldehydes, ketones, alcohols and formates were also quantified. -- Highlights: ► A specialist NMR rig was designed to measure the oxidation of FAME in situ. ► Oxidation of 18:1, 18:2 and 18:3 was observed over 24 h at 110 °C. ► The maximum rate was found at the start of the reaction for 18:3. ► The rate was highest for 18:2 after 300 min but never reached a maximum for 18:1.

  1. Virginia Water Central

    OpenAIRE

    Virginia Water Resources Research Center

    1999-01-01

    This newsletter features articles on water-related science, policy, and law. Distributed to state agency representatives, faculty, students and interested citizens, it aims to provide current information, statistics, news, and notices related to water resources in Virginia. Special Issue on Water Research

  2. Liquid metal reactor KALIMER development - Study on the high temperature properties of the steam generator tubing for LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Yong Soo; Kim, Soon Tae; Park, Hui Sang; Kim, Soo Han [Yonsei University, Seoul (Korea); Kim, Young Sik [Andong National University, Andong (Korea)

    1999-04-01

    This work dealt with the evaluation of super stainless steels for steam generator tubing of LMFBR. The experimental alloys were designed to simulate the elimination of alloying elements, in special, C and N. Regardless of carbon contents, super stainless steels showed the excellent properties (tensile properties and corrosion resistance) than those of 9Cr-1Mo steel. Nitrogen content has affected positively the ultimate tensile strength and yield strength by TT(Thermal Treatment), but the elongation was reduced by TT in case of nitrogen free alloy and the elongation was largely increased by TT in case of nitrogen bearing alloys. In acidic chloride environment, nitrogen has influenced a little on corrosion potential and critical current density, but largely on passive current density, especially, at high potential. However, the trend of corrosion potential and critical current density by nitrogen was similar to the results in acidic solutions, but passive current density was largely affected by nitrogen content of stainless steels. 29 refs., 24 figs., 8 tabs. (Author)

  3. Reactors

    International Nuclear Information System (INIS)

    Purpose: To provide a spray cooling structure wherein the steam phase in a bwr reactor vessel can sufficiently be cooled and the upper cap and flanges in the vessel can be cooled rapidly which kept from direct contaction with cold water. Constitution: An apertured shielding is provided in parallel spaced apart from the inner wall surface at the upper portion of a reactor vessel equipped with a spray nozzle, and the lower end of the shielding and the inner wall of the vessel are closed to each other so as to store the cooling water. Upon spray cooling, cooling water jetting out from the nozzle cools the vapor phase in the vessel and then hits against the shielding. Then the cooling water mostly falls as it is, while partially enters through the apertures to the back of the shielding plate, abuts against stoppers and falls down. The stoppers are formed in an inverted L shape so that the spray water may not in direct contaction with the inner wall of the vessel. (Horiuchi, T.)

  4. Fabrication of seamless calandria tubes

    International Nuclear Information System (INIS)

    Full text: Calandria tube is a large diameter, thin walled zircaloy-4 tube and is an important structural component of PHWR type of reactors. These tubes are lifetime components and remain during the full life of the reactor. Calandria tubes are classified as extremely thin walled tubes with a diameter to wall thickness ratio of around 96. Such thin walled tubes are conventionally produced by seam welded route comprising of extrusion of slabs followed by a series of hot and rolling passes, shaping into O-shape and eventual welding. An alternative and superior method of fabricating the calandria tubes, the seamless route, has been developed, which involves hot extrusion of mother blanks followed by three successive cold pilger reductions. Eccentricity correction of the extruded blanks is carried out on a special purpose grinding equipment to bring the wall thickness variation within permissible limits. Predominant wall thickness reductions are given during cold pilgering to ensure high Q-factor values. The texture in the finished tubes could be closely, controlled with an average fr value of 0.65. Pilgering parameters and tube guiding system have been specially designed to facilities rolling of thin walled tubes. Seamless calandria tubes have distinct advantages over welded tubes. In addition to the absence of weld, they are dimensionally more stable, lighter in weight and possess uniform grains with superior grain size. The cycle time from billet to finished product is substantially reduced and the product is amenable to high level of quality assurance. The most significant feature of the seamless route is its material recovery over welded route. Residual stresses measured in the tubes indicate that these are negligible and uniform along the length of the tube. In view of their superior quality, the first charge of seamless calandria tubes will be rolled into the first 500 MWe Pressurised Heavy Water Reactor at Tarapur

  5. Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors

    International Nuclear Information System (INIS)

    The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube No. 6. The sheets and forgings of Zircaloy 4 were examined in the temperature range between -256degC and 250degC. The thickness of the sheets are 3, 4, 5 and 10 mm, the maximum diameter of the forgings was 560 mm. This great forging diameters are not be treated in the ASTM rule B 351 for nuclear material, so a special approval with independent experts was necessary. The requirements for the material examinations were specified in a material specification and material test sheets which based on the ASTM rules B 351 and B 352 with additional restriction and additional requirements of the basic safety concept for nuclear power plants in Germany, which was taken into consideration in the nuclear licensing procedure. Charpy-V samples were carried out in the temperature range between -256degC and 150degC to get more information on the ductile behaviour of the Zircaloy 4. The results of the sheet examination confirm the requirements of the specifications, the results of the forging examination in the tangential testing direction are lower than specified and expected for the tensile strength. The axial and transverse values confirm the specification requirements. For the strength calculation of the pressure retaining wall a reduced material value for the forgings has to taken into consideration. The material behaviour of Zircaloy 4 under irradiation up to a fluence of ∼1·1022 n/cm2 was investigated. The loss of ductility was determined. As an additional criteria the variation of the fracture toughness was studies. Fracture mechanic calculations of the material were carried out in the licensing procedure with the focus to fulfill the leak before rupture criteria of the

  6. Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors

    International Nuclear Information System (INIS)

    The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube No. 6. The sheets and forgings of Zircaloy 4 were examined in the temperature range between -256 deg. C and 250 deg. C. The thickness of the sheets are 3, 4, 5 and 10 mm, the maximum diameter of the forgings was 560 mm. This great forging diameters are not be treated in the ASTM rule B 351 for nuclear material, so a special approval with independent experts was necessary. The requirements for the material examinations were specified in a material specification and material test sheets which based on the ASTM rules B 351 and B 352 with additional restriction and additional requirements of the basic safety concept for nuclear power plants in Germany, which was take into consideration in the nuclear licensing procedure. Charpy-V samples were carried out in the temperature range between -256 deg. C and 150 deg. C to get more information on the ductile behaviour of the Zircaloy 4. The results of the sheet examination confirm the requirements of the specifications, the results of the forging examination in the tangential testing direction are lower than specified and expected for the tensile strength. The axial and transverse values confirm the specification requirements. For the strength calculation of the pressure retaining wall a reduced material value for the forgings has to take into consideration. The material behaviour of Zircaloy 4 under irradiation up to a fluence of ∼ 1x1022 n/cm2 was investigated. The loss of ductility was determined. As an additional criteria the variation of the fracture toughness was studies. Fracture mechanic calculations of the material were carried out in the licensing procedure with the focus to fulfill the leak before rupture

  7. Developing a Topographic Model to Predict the Northern Hardwood Forest Type within Carolina Northern Flying Squirrel (Glaucomys sabrinus coloratus) Recovery Areas of the Southern Appalachians

    OpenAIRE

    Andrew Evans; Richard Odom; Lynn Resler; W. Mark Ford; Steve Prisley

    2014-01-01

    The northern hardwood forest type is an important habitat component for the endangered Carolina northern flying squirrel (CNFS; Glaucomys sabrinus coloratus) for den sites and corridor habitats between boreo-montane conifer patches foraging areas. Our study related terrain data to presence of northern hardwood forest type in the recovery areas of CNFS in the southern Appalachian Mountains of western North Carolina, eastern Tennessee, and southwestern Virginia. We recorded overstory species co...

  8. University to promote Virginia Cycling and Pedestrian Awareness Week

    OpenAIRE

    West, Hilary

    2009-01-01

    Virginia Tech has joined DRIVE SMART Virginia, Bike Walk Virginia, the Virginia Highway Safety Office of the Commonwealth's Department of Motor Vehicles to promote Virginia Cycling and Pedestrian Awareness Week scheduled for Sept. 13-20.

  9. Pressure tube life management in CANDU-6 nuclear plant

    International Nuclear Information System (INIS)

    Operating parameters of pressure tube in CANDU-6 reactor, the relation between pressure tube life and plant life improvement of pressure tube by AECL in past years were summarized, and the factors affecting pressure tube life, idea and main measures of pressure tube life management in QINSHAN CANDU-6 power plant introduced

  10. Pathology of aural abscesses in free-living Eastern box turtles (Terrapene carolina carolina).

    Science.gov (United States)

    Brown, Justin D; Richards, Jean M; Robertson, John; Holladay, Steven; Sleeman, Jonathan M

    2004-10-01

    Aural abscess or abscess of the middle ear is common in free-living Eastern box turtles (Terrapene carolina carolina) of Virginia (USA) and elsewhere. Although its etiology remains unknown, hypovitaminosis A has been suggested on the basis of similar lesions occurring in captive chelonians fed diets that are deficient in vitamin A. This hypothesis was supported by significantly greater body burdens of organochlorine compounds (reported disruptors of vitamin A metabolism) and a nonsignificant trend toward lower serum and hepatic vitamin A levels in free-living box turtles with this lesion. The tympanic epithelium was evaluated in 27 box turtles (10 with aural abscesses and 17 without). Lesions of the tympanic epithelium of box turtles with aural abscesses included hyperplasia, squamous metaplasia, hyperemia, cellular sloughing, granulomatous inflammation, and bacterial infection. These changes were more severe in turtles with aural abscesses than in those without and were more severe in tympanic cavities that had an abscess compared to those without when the lesion was unilateral. Organs from 21 box turtles (10 with aural abscesses and 11 without) from the study population were examined for microscopic lesions, and minimal histopathologic changes were found, none of which were similar to those found in the tympanic epithelium. Histopathologic changes in box turtles with aural abscesses were consistent with a syndrome that may involve hypovitaminosis A. PMID:15650088

  11. Phaeohyphomycosis in a free-living eastern box turtle (Terrapene carolina carolina).

    Science.gov (United States)

    Joyner, Priscilla H; Shreve, Allison A; Spahr, John; Fountain, Andrea L; Sleeman, Jonathan M

    2006-10-01

    A free-ranging eastern box turtle (Terrapene carolina carolina) was referred to the Wildlife Center of Virginia with a three-month history of marked swelling of the right hind limb initially diagnosed as chromomycosis by histopathology. Hematology revealed severe anemia (9%), leukocytosis (12.8 cells x 10(3)/microl), heterophilia (6.14 cells x 10(3)/microl), and monocytosis (0.51 cells x 10(3)/microl). Gross necropsy revealed a firm, encapsulated 3 x 1 cm subcutaneous mass filled with dark brown-black, friable necrotic material of the distal right hind limb. Microscopically, the mass was characterized by a granulomatous inflammatory process with numerous multinucleated histiocytic giant cells. Fungal elements were present within necrotic centers and associated with multinucleated cells. Special stains revealed numerous phaeoid hyphae and yeast; Exophiala jeanselmei was isolated by routine mycologic culture. Phaeohyphomycosis was diagnosed based on the histologic appearance of the fungal elements within the mass and culture results. There was no histopathological evidence of systemic infection. This is the first report of phaeohyphomycosis caused by fungi of the genus Exophiala in free-living reptiles. PMID:17255461

  12. Analysis of steam-generator tube-rupture events combined with auxiliary-feedwater control-system failure for Three Mile Island-Unit 1 and Zion-Unit 1 pressurized water reactors

    International Nuclear Information System (INIS)

    A steam-generator tube-rupture (SGTR) event combined with loss of all offsite alternating-current power and failure of the auxiliary-feedwater (AFW) control system has been investigated for the Three Mile Island-Unit 1 (TMI-1) and Zion-Unit 1 (Zion-1) pressurized water reactors. The Transient Reactor Analysis Code was used to simulate the accident sequence for each plant. The objectives of the study were to predict the plant transient response with respect to tube-rupture flow termination, extent of steam generator overfill, and thermal-hydraulic conditions in the steam lines. Two transient cases were calculated: (1) a TMI-1 SGTR and runaway-AFW transient, and (2) a Zion-1 SGTR and runaway-AFW transient. Operator actions terminated the tube-rupture flow by 1342 s (22.4 min) and 1440 s (24.0 min) for TMI-1 and Zion-1, respectively, but AFW injection was continued. The damaged steam generator (DSG) overfilled by 1273 s (21.2 min) for the TMI-1 calculation and by 1604 s (26.7 min) for the Zion-1 calculation. The DSG steam lines were completely filled by 1500 s (25 min) and 2000 s (33.3 min) for TMI-1 and Zion-1, respectively. The maximum subcooling in the steam lines was approx.63 K (approx.1130F) for TMI-1 and approx.44 K (approx.800F) for Zion-1

  13. Study of the influence of pH value and passivation in nitrate solutions on corrosion resistance of reactor BN-600 cladding tubes of steel EhP-450

    International Nuclear Information System (INIS)

    A study is made into the influence of steel EhP-450 cladding tube operational conditions in BN-600 reactor, a pH value of fuel storage pool water medium and the passivation in iron and chromium nitrate aqueous solutions on corrosion resistance of the steel. The corrosion resistance is judged from the mass loss. It is established that at inner surfaces of cladding tubes the chromium-depleted and carbon-enriched layers of reduced corrosion resistance are formed in operation. Analytical expressions are obtained which allow to have quantitative correlations between corrosion characteristics (corrosion rate, corrosion products entrainment, surface layer thickness) and operational conditions for steel EhP-450 in BN-600 reactor. The studies provide support for the view that the passivation treatment of EhP-450 cladding tubes in a Fe(NO3)3 solution with the aim of their corrosion protection when holding in the water at pH ≥ 8.0 in a fuel storage pool has considerable promise

  14. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    The description is given of a water cooled nuclear reactor comprising a core, cooling water that rises through the core, vertical guide tubes located inside the core and control rods vertically mobile in the guide tubes. In this reactor the cooling water is divided into a first part introduced at the bottom end of the core and rising through it and a second part introduced at the top end of the guide tubes so as to drop in them

  15. Tube-in-shell heat exchangers

    International Nuclear Information System (INIS)

    Tube-in-shell heat exchangers normally comprise a bundle of parallel tubes within a shell container, with a fluid arranged to flow through the tubes in heat exchange with a second fluid flowing through the shell. The tubes are usually end supported by the tube plates that separate the two fluids, and in use the tube attachments to the tube plates and the tube plates can be subject to severe stress by thermal shock and frequent inspection and servicing are required. Where the heat exchangers are immersed in a coolant such as liquid Na such inspection is difficult. In the arrangement described a longitudinally extending central tube is provided incorporating axially spaced cylindrical tube plates to which the opposite ends of the tubes are attached. Within this tube there is a tubular baffle that slidably seals against the wall of the tube between the cylindrical tube plates to define two co-axial flow ducts. These ducts are interconnected at the closed end of the tube by the heat exchange tubes and the baffle comprises inner and outer spaced walls with the interspace containing Ar. The baffle is easily removable and can be withdrawn to enable insertion of equipment for inspecting the wall of the tube and tube attachments and to facilitate plugging of defective tubes. Cylindrical tube plates are believed to be superior for carrying pressure loads and resisting the effects of thermal shock. Some protection against thermal shock can be effected by arranging that the secondary heat exchange fluid is on the tube side, and by providing a thermal baffle to prevent direct impingement of hot primary fluid on to the cylindrical tube plates. The inner wall of the tubular baffle may have flexible expansible region. Some nuclear reactor constructions incorporating such an arrangement are described, including liquid metal reactors. (U.K.)

  16. Virginia Tech graduate student chosen to be Virginia Governor's Fellow

    OpenAIRE

    Harris, Sally L.

    2005-01-01

    Lindsay Potts produced for the state Secretary of Agriculture and Forestry a study of the feasibility of producing and using biodiesel fuel as part of her work with the Virginia Governor's Fellows Program.

  17. Flock sizes and sex ratios of canvasbacks in Chesapeake Bay and North Carolina

    Science.gov (United States)

    Haramis, G.M.; Derleth, E.L.; Link, W.A.

    1994-01-01

    Knowledge of the distribution, size, and sex ratios of flocks of wintering canvasbacks (Aythya valisineria) is fundamental to understanding the species' winter ecology and providing guidelines for management. Consequently, in winter 1986-87, we conducted 4 monthly aerial photographic surveys to investigate temporal changes in distribution, size, and sex ratios of canvasback flocks in traditional wintering areas of Chesapeake Bay and coastal North Carolina. Surveys yielded 35mm imagery of 194,664 canvasbacks in 842 flocks. Models revealed monthly patterns of flock size in North Carolina and Virginia, but no pattern of change in Maryland. A stepwise analysis of flock size and sex ratio fit a common positive slope (increasing proportion male) for all state-month datasets, except for North Carolina in February where the slope was larger (P lt 0.001). State and month effects on intercepts were significant (P lt 0.001) and confirmed a previously identified latitudinal gradient in sex ratio in the survey region. There was no relationship between flock purity (% canvasbacks vs. other species) and flock size except in North Carolina in January, February, and March when flock purity was related to flock size. Contrasting characteristics in North Carolina with regard to flock size (larger flocks) and flock purity suggested that proximate factors were reinforcing flocking behavior and possibly species fidelity there. Of possible factors, the need to locate foraging sites within this large, open-water environment was hypothesized to be of primary importance. Comparison of January 1981 and 1987 sex ratios indicated no change in Maryland, but lower (P lt 0.05) canvasback sex ratios (proportion male) in Virginia and North Carolina.

  18. Virginia ESI: REPTPT (Reptile Points)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains sensitive biological resource data for sea turtles in Virginia. Vector points in this data set represent nesting sites. Species-specific...

  19. Virginia ESI: REPTILES (Reptile Polygons)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains sensitive biological resource data for sea turtles and estuarine turtles in Virginia. Vector polygons in this data set represent turtle...

  20. Virginia ESI: INVERT (Invertebrate Polygons)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains sensitive biological resource data for marine, estuarine, and rare invertebrate species in Virginia. Vector polygons in this data set...

  1. Virginia ESI: Wetlands (Wetland Polygons)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains vector polygons representing the coastal wetlands for Virginia, classified according to the Environmental Sensitivity Index (ESI)...

  2. Virginia ESI: FISH (Fish Polygons)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains sensitive biological resource data for marine, estuarine, anadromous, and brackishwater fish species in Virginia. Vector polygons in this...

  3. Current practices for ultrasonic and radiographic examination of tubes, tube plates and tube-plate welds of tube bundles in heat exchangers. Chapter 3

    International Nuclear Information System (INIS)

    The chapter describes the ultrasonic and radiographic inspection procedures that are applied to heat exchanger tube bundles. The inspection process starts with the ultrasonic examination of the tubes and tube plates during manufacture, followed by radiography of the tube-to-tube-plate welds during fabrication of the tube bundle. Ultrasonic methods are explained for welds which are amenable to this type of inspection. For the in-service inspection of tube bundles the chapter relates the authors' experiences on the ultrasonic inspection of tubes and tube plates in the Prototype Fast Breeder Reactor at Dounreay. At the end of the chapter some comments are made about future ultrasonic and radiographic developments for tube bundles. (author)

  4. Virginia Agritourism: A Profitability Analysis

    OpenAIRE

    Lucha, Christopher Ryan

    2014-01-01

    Agritourism in Virginia is a rapidly growing industry that adds additional income to a farming operation, and helps mitigate risk. Therefore it has become a good strategy for farmers to generate higher levels of profit, but much of the literature in Virginia and surrounding states focuses more on the motivations of operators for starting their agritourism venture. Thus, the purpose of this paper is to empirically analyze demographic, operational, and financial factors and evaluate their cor...

  5. The Analysis of Virginia Woolf

    Institute of Scientific and Technical Information of China (English)

    戴寒路

    2014-01-01

    <正>Virginia Woolf is a British writer,one of the representatives criticism at home,stream of consciousness novel."Mark on the Wall"is her first article typical stream of consciousness works.Virginia Woolf has a complicated family background.The nine homes,two groups of children between age and personality clashes often occur some contradictions and conflicts.Woolf’s half-brother and two of her injuries left her permanently

  6. Carolinas Energy Career Center

    Energy Technology Data Exchange (ETDEWEB)

    Classens, Anver; Hooper, Dick; Johnson, Bruce

    2013-03-31

    Central Piedmont Community College (CPCC), located in Charlotte, North Carolina, established the Carolinas Energy Career Center (Center) - a comprehensive training entity to meet the dynamic needs of the Charlotte region's energy workforce. The Center provides training for high-demand careers in both conventional energy (fossil) and renewable energy (nuclear and solar technologies/energy efficiency). CPCC completed four tasks that will position the Center as a leading resource for energy career training in the Southeast: • Development and Pilot of a New Advanced Welding Curriculum, • Program Enhancement of Non-Destructive Examination (NDE) Technology, • Student Support through implementation of a model targeted toward Energy and STEM Careers to support student learning, • Project Management and Reporting. As a result of DOE funding support, CPCC achieved the following outcomes: • Increased capacity to serve and train students in emerging energy industry careers; • Developed new courses and curricula to support emerging energy industry careers; • Established new training/laboratory resources; • Generated a pool of highly qualified, technically skilled workers to support the growing energy industry sector.

  7. Infectious uveitis in Virginia

    Directory of Open Access Journals (Sweden)

    Engelhard SB

    2015-08-01

    Full Text Available Stephanie B Engelhard,1 Zeina Haddad,1 Asima Bajwa,1 James Patrie,2 Wenjun Xin,2 Ashvini K Reddy1 1Department of Ophthalmology, 2Department of Public Health Sciences, University of Virginia, Charlottesville, VA, USA Purpose: To report the causes, clinical features, and outcomes of infectious uveitis in patients managed in a mid-Atlantic tertiary care center.Methods: Retrospective, observational study of infectious uveitis patients seen at the University of Virginia from 1984 to 2014.Results: Seventy-seven of 491 patients (15.7% were diagnosed with infectious uveitis (mean age 58 years, 71.4% female, 76.6% Caucasian. The mean follow-up was 5 years. Anterior uveitis was the most common anatomic classification (39 patients, 50.6% followed by panuveitis (20 patients, 26.0% and posterior uveitis (18 patients, 23.4%. The most common infectious etiology was herpetic anterior uveitis (37 patients, 48.1% followed by toxoplasma uveitis (14 patients, 18.2%. The most prevalent viral pathogen was varicella-zoster virus (21 patients, 27.3% followed by herpes simplex virus (20 patients, 26.0%. Acute retinal necrosis (ARN was diagnosed in 14 patients (18.2%. Aqueous humor yielded an etiologic diagnosis in seven (50% of ARN patients, four of whom tested positive for cytomegalovirus and three for varicella-zoster virus. On presentation, 43 patients (55.8% had a visual acuity (VA better than 20/40 and 17 (22.1% had a VA worse than 20/200. VA at the final follow-up was better than 20/40 in 39 patients (50.6% and worse than 20/200 in 22 patients (28.6%. In all, 16 (20.8% and 10 (13.0% patients required cataract and vitrectomy surgery, respectively. A total of 14 patients (18.2% were on glaucoma topical treatment and four (5.2% required glaucoma surgery.Conclusion: The most common type of infectious uveitis seen over the study period was herpetic anterior uveitis secondary to varicella-zoster virus or herpes simplex virus, found to be most prevalent in patients

  8. Sealing device for nuclear power reactor

    International Nuclear Information System (INIS)

    The sealing device is to stop a leak on a reactor pressure vessel where control of the output of reactor is arranged by control rods which are handled by drives connected to control rods and bars in tubes which penetrate the reactor wall. Each tube has a supporting case on the inside of the wall opened to the hole and welded to the tube. The weld may crack and leak. Then an inner sealing tube made of soft metallic material whose outer surface is conical is drawn on to the tube over which an outer sealing tube made of hard metallic material and conical inner surface is placed. On both sides of the crack special adhering planes are formed between the inner sealing tube and the tubes or the supporting case. When the outer sealing tube is pressed over the inner sealing tube, the conical surfaces tighten it against the tube and the supporting case

  9. Update on seamless calandria tube development and qualification

    International Nuclear Information System (INIS)

    AECL is undertaking the qualification of the production of seamless calandria tubes as replacement components for installation in reactors during retubing. Seamless tube prototypes made from Zircaloy-2 possessing a suitable crystallographic texture have been shown to be significantly stronger than seam-welded tubes under both rising pressure and sustained pressure conditions in a simulated reactor loading. This paper describes the seamless calandria tube development program and current status. (author)

  10. Whither North Carolina furniture manufacturing?

    OpenAIRE

    Robert L. Lacy

    2004-01-01

    North Carolina's furniture manufacturing industry has contracted in recent years as imports have gained a greater share of the domestic furniture market. Rapid growth of the furniture industry in China and a surge in exports from that country to the United States in particular have contributed to plant closings and consolidation of operations in the state. North Carolina's furniture manufacturers are adapting to the emergence of global competition and are developing new corporate strategies t...

  11. Status of R and D Activities related to Axial and Radial Creep of Pressure tubes of Heavy Water Reactors in India

    International Nuclear Information System (INIS)

    Scope of presentation: • R and D Strength of the organisation; • Brief introduction about PHWRs in India – Operating and under construction; • Issues associated with axial elongation and radial expansion due to neutron enhanced creep in pressure tube; • Status of work done till date; • Important Points related to present CRP

  12. Virginia Tech receives NIH funds brucellosis research

    OpenAIRE

    Trulove, Susan

    2004-01-01

    Virginia Tech researchers from Virginia Bioinformatics Institute (VBI) and Virginia-Maryland Regional College of Veterinary Medicine (VMRCVM) have received $300,000 from the National Institutes of Health (NIH) to study brucellosis. Caused by Brucella bacteria, a potential bioterrorism agent, brucellosis is common to animals and some strains infect humans.

  13. Virginia Regional Seismic Network. Final report (1986--1992)

    Energy Technology Data Exchange (ETDEWEB)

    Bollinger, G.A.; Sibol, M.S.; Chapman, M.C.; Snoke, J.A. [Virginia Polytechnic Inst. and State Univ., Blacksburg, VA (US). Seismological Observatory

    1993-07-01

    In 1986, the Virginia Regional Seismic Network was one of the few fully calibrated digital seismic networks in the United States. Continued operation has resulted in the archival of signals from 2,000+ local, regional and teleseismic sources. Seismotectonic studies of the central Virginia seismic zone showed the activity in the western part to be related to a large antiformal structure while seismicity in the eastern portion is associated spatially with dike swarms. The eastern Tennessee seismic zone extends over a 300x50 km area and is the result of a compressive stress field acting at the intersection between two large crustal blocks. Hydroseismicity, which proposes a significant role for meteoric water in intraplate seismogenesis, found support in the observation of common cyclicities between streamflow and earthquake strain data. Seismic hazard studies have provided the following results: (1) Damage areas in the eastern United States are three to five times larger than those observed in the west. (2) Judged solely on the basis of cataloged earthquake recurrence rates, the next major shock in the southeast region will probably occur outside the Charleston, South Carolina area. (3) Investigations yielded necessary hazard parameters (for example, maximum magnitudes) for several sites in the southeast. Basic to these investigations was the development and maintenance of several seismological data bases.

  14. Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

    International Nuclear Information System (INIS)

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source

  15. Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    2000-07-01

    The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source.

  16. Design, fabrication, and testing of gadolinium-shielded metal fuel samples in the hydraulic tube of the high flux isotope reactor

    International Nuclear Information System (INIS)

    The use of hydraulic rabbit capsules inserted into and ejected from the core of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) during full power operation allows for precise control of the neutron fluence in fueled experiments. Rabbit capsules with strong thermal neutron absorbers must be used to screen out thermal neutrons, thereby reducing the heat generation rate while maintaining the fast neutron flux that produces displacement damage similar to fast reactor type conditions. However, rapid insertion and ejection of rabbit capsules containing a strong neutron absorber causes a reactivity response in the reactor that has the potential to engage the HFIR safety response system which could result in an unplanned shutdown. Therefore, a set of tests were performed to provide the data needed to establish limits on the reactivity worth that can be ejected from the hydraulic facility without causing a reactor shutdown. This paper will describe the design, operation, and results of the reactivity measurements undertaken to understand the reactor response to insertion of the gadolinium-lined rabbit capsules. (author)

  17. LOGGERHEAD SEA TURTLE LATE NESTING ECOLOGY IN VIRGINIA BEACH, VIRGINIA

    Science.gov (United States)

    T'he.loggerhead sea turtle (Caretta came is the only recurrent nesting species of sea turtle in southeastern Virginia (Lutcavage & Musick, 1985; Dodd, 1988). Inasmuch as the loggerhead is a federally threatened species, the opportunity to gather data on its nesting ecology is imp...

  18. Modeling of the anisotropic elasto-viscoplastic behavior of irradiated cladding tubes between zero and four cycles in power water reactor

    International Nuclear Information System (INIS)

    As more than 75 % of electricity in France is generated by nuclear energy, the French PWR have an important role in the grid regulation. Load follow has become a normal operating mode. In addition, EDF wants to increase the discharge burn-up of the fuel sub-assemblies to about 60 GWd/tU. Those operating conditions lead to numerous pellet cladding interaction (PCI) transients. In order to improve the accuracy of the stress-strain level in the cladding, a model to describe the anisotropic viscoplastic behavior of irradiated cold work stress relieved Zircaloy-4 cladding tubes has been developed. This model is presented in the present report. The anisotropy is described in the modelling through fourth orders' tensors affecting the flow direction, the linear part, as well as the dynamic and static recoveries of the kinematical hardening variables. That model has been identified at 350 deg C and 400 deg C on cladding tubes representative of the industrial products introduced in the PWR of EDF. Temperature effects affect the state equation and the static recovery term of the model. The ability of the model to simulate the unirradiated state has been shown on complex loadings representative of a PCI loading at 350, 380 and 400 deg C. The study shows an irradiation hardening of the cladding tubes and a decrease in creep deformation. These changes in mechanical properties are significant until 45.1024 n/m2 (E > 1 MeV). After this fluence level, the evolution of mechanical behaviour is not so pronounced. No change in the elastic behavior of the tested tubes has been notified. A state variable of damage function of the fluence imposes an increase of the kinematical hardening and a decrease of the static recovery. This formulation is identified on monotonous tensile and creep tests performed at 350 deg C on irradiated cladding tubes, and validated at 380 and 400 deg C. The application of the modeling is illustrated by a finite element calculation of an irradiated cladding tubes

  19. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

  20. Numerical Analysis on the Calandria Tubes in the Moderator of a Heavy Water Reactor Using OpenFOAM and Other Codes

    Science.gov (United States)

    Chang, Se-Myong; Kim, Hyoung Tae

    2014-06-01

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors.

  1. Numerical analysis on the calandria tubes in the moderator of a heavy water reactor using OpenFOAM and other codes

    International Nuclear Information System (INIS)

    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors. (authors)

  2. Analysis of the effect of tube arrangement and inclination on pressure drop in an intermediate heat exchanger of liquid metal reactor

    Energy Technology Data Exchange (ETDEWEB)

    ChoiI, Seok Ki; Choi, Il Kon; Nam, Ho Yun; Choi, Jong Hyeun [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-05-01

    An experimental study on the effect of tube arrangement and inclination on the pressure drop in the intermediate heat exchanger is performed. Measurements are made for pressure drop in the triangular and rotated triangular tue arrays whose inclined angles are 30, 45, 60, 75 and 90 degrees. The pitch to tube diameter ratio is 1.6 and the range of Reynolds number based on the free stream velocity and tube diameter is 870-64,000. The experimental results show that the magnitude of dimensionless pressure drop increases with the inclined angle and decreases significantly when the inclined angle is less than 45 degree. The previous correlations are evaluated using the experimental data. The ESDU correlation agrees well with the present data for the triangular arrays. But some discrepancies are observed for the rotated triangular arrays when the inclined angles are 45 and 30 degrees. The Idel'chik correlation generally agrees well with the measured data for the rotated triangular arrays except for inclined angle of 30 degree. The Idel'chik correlation needs modification for the triangular arrays. The modified Idel'chik correlation agrees well with the measure data within 10%. 32 refs., 59 figs., 11 tabs. (Author)

  3. Thermal or epithermal reactor

    International Nuclear Information System (INIS)

    In a thermal or epithermal heavy-water reactor of the pressure tube design the reactivity is to be increased by different means: replacement of the moderator by additional rods with heavy metal in the core or in the reflector; separation of the moderator (heavy water) from the coolant (light water) by means of shroud tubes. In light-water reactor types neutron losses are to be influenced by using the heavy elements in different configurations. (orig./PW)

  4. Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors results of forming the material and welding processes

    International Nuclear Information System (INIS)

    The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube no. 6. For fabrication of the vessels and piping parts it was necessary to form the base material and calibrate the sheets or welded parts with necessary heat treatments. Additional to the technical specifications preliminary material investigations and production test of welded and unwelded material were carried out of the formed parts up to a cold work of 5%. Further one with respect to the material thickness of 3, 4, 5 and 10 mm of the used sheets, welding procedure test before the fabrication and welding production tests during fabrication were carried out of the base material combination sheet/sheet and sheet/forging. Electronic beam welding was used for the welding process. Material tests as tensile tests, charpy-V-tests, bend tests, metallographic tests, hardness tests, radiographic tests a.s.o. were carried out. The results of the examinations confirm the specified requirements. For the material forming process an optimization was necessary after the preliminary results to get final sufficient material behaviour results. (orig.)

  5. Alternate tube plugging criteria for steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Cueto-Felgueroso, C.; Aparicio, C.B. [Tecnatom, S.A., Madrid (Spain)

    1997-02-01

    The tubing of the Steam Generators constitutes more than half of the reactor coolant pressure boundary. Specific requirements governing the maintenance of steam generator tubes integrity are set in Plant Technical Specifications and in Section XI of the ASME Boiler and Pressure Vessel Code. The operating experience of Steam Generator tubes of PWR plants has shown the existence of some types of degradatory processes. Every one of these has an specific cause and affects one or more zones of the tubes. In the case of Spanish Power Plants, and depending on the particular Plant considered, they should be mentioned the Primary Water Stress Corrosion Cracking (PWSCC) at the roll transition zone (RTZ), the Outside Diameter Stress Corrosion Cracking (ODSCC) at the Tube Support Plate (TSP) intersections and the fretting with the Anti-Vibration Bars (AVBs) or with the Support Plates in the preheater zone. The In-Service Inspections by Eddy Currents constitutes the standard method for assuring the SG tubes integrity and they permit the monitoring of the defects during the service life of the plant. When the degradation reaches a determined limit, called the plugging limit, the SG tube must be either repaired or retired from service by plugging. Customarily, the plugging limit is related to the depth of the defect. Such depth is typically 40% of the wall thickness of the tube and is applicable to any type of defect in the tube. In its origin, that limit was established for tubes thinned by wastage, which was the predominant degradation in the seventies. The application of this criterion for axial crack-like defects, as, for instance, those due to PWSCC in the roll transition zone, has lead to an excessive and unnecessary number of tubes being plugged. This has lead to the development of defect specific plugging criteria. Examples of the application of such criteria are discussed in the article.

  6. Analysis of autofrettaged metal tubes

    International Nuclear Information System (INIS)

    Thick-walled cylinders are widely used as compressor cylinders, pump cylinders, high pressure tubing, process reactors and vessels, nuclear reactors, isostatic vessels and gun barrels. In practice, cylinders are generally subjected to sudden and frequently drastic pressure fluctuations, such as the pressure generated in a gun barrel upon the firing of the weapon, pressure reversals in pump cylinders or in process reactors employing high-pressure piping, necessitating enhanced strength of such cylinders. A process for enhancing the strength of thick-walled cylinders has been in service, and is referred to as 'autofrettage'. It extends the service life of the cylinder. The autofrettage is achieved by increasing elastic strength of a cylinder with various methods such as hydraulic pressurization, mechanical swaging, or by utilizing the pressure of a powder gas. This research work deals with the hydraulic and mechanical autofrettage of metal tubes with the objective to attain enhanced strength. Five metal tubes are taken randomly for analysis purpose. The experimental data for five metal tubes is obtained to analyze the behavior of different parameters used during, before, and after autofrettage process. For this research, two-stage autofrettage is taken into consideration. The modeling of the metal tube is carried out in WildFire-ProEngineering, and for analysis purpose, finite element software ANSYS7 and COSMOS are used. The graphical analysis of swage autofrettage is carried out using MATLAB7. The results are validated using available experimental and numerical data. (author)

  7. On the distribution of temperatures in steam generator tubes at tube support plate Intersections

    International Nuclear Information System (INIS)

    This analysis was initiated to examine the temperature fields in the steam generator tube in the vicinity of the tube support plates. It is assumed that the flow of the secondary coolant is severely disturbed there, which causes local heating of the tube surface. Different designs of tube support plates (a drilled hole - NE Krsko, broached trefoil and broached quatrefoil designs) were assessed and compared. Inside the drilled hole tube support plate, the temperature of the reactor coolant. Inside broached trefoil and quatrefoil support plates, the tube surface temperature reaches about 10K less than reactor coolant temperature. The most important result concerning the Krsko specific conditions is that the frequency of the detected defects can be correlated with the temperature of the tube outer surface and void fraction of the secondary coolant. (author)

  8. Tube-in-shell heat exchangers

    International Nuclear Information System (INIS)

    A tube-in-shell heat exchanger is described for use in liquid metal cooled fast breeder reactor constructions. The system consists of a bundle of heat exchange tubes with a central spine extending longitudinally through the shell and a series of longitudinally spaced transverse grids resiliently mounted on the central spine within the shell to provide transverse support for bracing the tubes apart. (U.K.)

  9. Gastrostomy Tube (G-Tube)

    Science.gov (United States)

    ... endoscope (a thin, flexible tube with a tiny camera and light at the tip) inserted through the ... Nemours Foundation, iStock, Getty Images, Corbis, Veer, Science Photo Library, Science Source Images, Shutterstock, and Clipart.com

  10. Ear tube insertion

    Science.gov (United States)

    Myringotomy; Tympanostomy; Ear tube surgery; Pressure equalization tubes; Ventilating tubes; Ear infection - tubes; Otitis - tubes ... trapped fluid can flow out of the middle ear. This prevents hearing loss and reduces the risk ...

  11. 75 FR 80547 - Carolina Power & Light Company, Shearon Harris Nuclear Power Plant, Unit No. 1; Exemption

    Science.gov (United States)

    2010-12-22

    ... COMMISSION Carolina Power & Light Company, Shearon Harris Nuclear Power Plant, Unit No. 1; Exemption 1.0... License No. NPF-63, which authorizes operation of the Shearon Harris Nuclear Power Plant (HNP), Unit 1... nuclear power reactors against radiological sabotage,'' published as a final rule in the Federal...

  12. Classification of West Virginia and Virginia Balsam Fir Communities Affected by Balsam Woolly Adelgid Infestation

    Data.gov (United States)

    US Fish and Wildlife Service, Department of the Interior — Balsam fir reaches its southern limit in several high elevation wetlands in West Virginia and highelevation forests and rock outcrops in Virginia. This study...

  13. 76 FR 54189 - Television Broadcasting Services; Hampton-Norfolk, Virginia; Norfolk, Virginia-Elizabeth City, NC

    Science.gov (United States)

    2011-08-31

    ... COMMISSION 47 CFR Part 73 Television Broadcasting Services; Hampton-Norfolk, Virginia; Norfolk, Virginia... 73 Television, Television broadcasting. Federal Communications Commission. Barbara A. Kreisman, Chief... Association (``HRETA''), the licensee of noncommercial educational television station WHRO-TV, channel...

  14. Nuclear reactor spacer assembly

    International Nuclear Information System (INIS)

    A fuel assembly for a nuclear reactor is disclosed wherein the fuel element receiving and supporting grid is comprised of a first metal, the guide tubes which pass through the grid assembly are comprised of a second metal and the grid is supported on the guide tubes by means of expanded sleeves located intermediate the grid and guide tubes. The fuel assembly is fabricated by inserting the sleeves, of initial outer diameter commensurate with the guide tube outer diameters, through the holes in the grid assembly provided for the guide tubes and thereafter expanding the sleeves radially outwardly along their entire length such that the guide tubes can subsequently be passed through the sleeves. The step of radial expansion, as a result of windows provided in the sleeves having dimensions commensurate with the geometry of the grid, mechanically captures the grid and simultaneously preloads the sleeve against the grid whereby relative motion between the grid and guide tube will be precluded

  15. Virginia Power's regulatory reduction program

    International Nuclear Information System (INIS)

    Virginia Power has two nuclear plants, North Anna and Surry Power Stations, which have two units each for a total of four nuclear units. In 1992, the Nuclear Regulatory Commission solicited comments from the nuclear industry to obtain their ideas for reducing the regulatory burden on nuclear facilities. Pursuant to the new regulatory climate, Virginia Power developed an internal program to evaluate and assess the regulatory and self-imposed requirements to which they were committed, and to pursue regulatory relief or internal changes where possible and appropriate. The criteria were that public safety must be maintained, and savings must be significant. Up to the date of the conference, over US$22 million of one-time saving had been effected, and US$2.75 million in annual savings

  16. Coiled tubing

    International Nuclear Information System (INIS)

    Oil and gas wells that flow on initial completion eventually reach a condition of liquid loading that kills the wells. This results form declining reservoir pressure, decreased gas volume (velocity), increased water production and other factors that cause liquids to accumulate at the bottom of the well and exert back pressure on the formation. This restricts or in some cases prevents fluid entry into the wellbore form the formation. Flowing production can be restored or increased by reducing surface backpressure, well bore stimulation, pressure maintenance or by installing a string of smaller diameter tubing. This paper reports on installation (hanging off) of a concentric string of coiled tubing inside existing production tubing which is an economically viable, safe, convenient and effective alterative for returning some of these liquid loaded )logged-up) wells to flowing status

  17. Evaluation of corrosion of 800GN alloy tubes in similar ambient to the secondary circuit of PWR reactor at 80 deg C

    International Nuclear Information System (INIS)

    In this work we investigated the effect of the presence of chloride ions (concentrations of 10, 50 and 250 ppb) and sulfate in a ratio of 1: 1 in a corrosion behavior of 800NG alloy tube to 80 ° C, in electrochemical cell to three electrodes. Such concentrations correspond to action levels used in the pipeline safety. Experiments in potential open circuit and cyclic polarization were used to characterize the corrosion behavior of the material. Morphological analysis of corrosion and corrosion products was performed by optical microscopy, scanning electron microscopy and X-ray diffraction. The results showed that the morphology of the attack located in a alloy 800GN is related to the ratio between the concentrations of chloride and sulfate ions in the medium

  18. Virginia Tech Corps of Cadets alumnus Matthew Van Arsdale named Virginia game Hokie Hero

    OpenAIRE

    Cox, Carrie

    2010-01-01

    Virginia Tech Corps of Cadets alumnus 2nd Lt. Matthew Van Arsdale, U.S. Army, who earned a degree in history from the College of Liberal Arts and Human Sciences and a minor in leadership studies from the Virginia Tech Corps of Cadets Rice Center for Leader Development in 2009 has been selected as the Hokie Hero for the Virginia Tech versus University of Virginia game.

  19. THE CAROLINA LUPUS STUDY (CLU)

    Science.gov (United States)

    Carolina Lupus (CLU) Study, an epidemiologic study of risk factors for systemic lupus erythematosus (SLE). SLE is a severe, chronic, systemic autoimmune disease that disproportionately affects women and African-Americans. The CLU Study focuses on measures of endogenous hormone ex...

  20. Assessment and management of ageing of major nuclear power plant components important to safety: CANDU pressure tubes

    International Nuclear Information System (INIS)

    The report documents the current practices for assessment and management of the ageing of the pressure tubes in CANDU reactors and Indian PHWTRs. Chapter headings are: fuel channel and pressure tube description, design basis for the fuel channel and pressure tube, degradation mechanisms and ageing concerns for pressure tubes, inspection and monitoring methods for pressure tubes,assessment methods and fitness-for-service guidelines for pressure tubes, mitigation methods for pressure tubes, and pressure tube ageing management programme

  1. Nuclear reactor repairing device

    International Nuclear Information System (INIS)

    Purpose: To enable free repairing of an arbitrary position in an LMFBR reactor. Constitution: A laser light emitted from a laser oscillator installed out of a nuclear reactor is guided into a portion to be repaired in the reactor by using a reflecting mirror, thereby welding or cutting it. The guidance of the laser out of the reactor into the reactor is performed by an extension tube depending into a through hole of a rotary plug, and the guidance of the laser light into a portion to be repaired is performed by the transmitting and condensing action of the reflecting mirror. (Kamimura, M.)

  2. Virginia Tech Featured On Science Coalition's website

    OpenAIRE

    Trulove, Susan

    2004-01-01

    Research at Virginia Tech that impacts everyday life will be featured May 16-22 on the Science Coalition's website, a comprehensive resource for information on federally funded science research. The Science Coalition website featuring Virginia Tech will present research by faculty members and students looking at obesity and stiff arteries, wireless networks and smart garments, brain tumors, cultural and archeological history of Southside Virginia, recovering discarded coal, theater production...

  3. Virginia Tech Mobile delivers news to cell phones, PDAs

    OpenAIRE

    Lazenby, Jenna

    2006-01-01

    Virginia Tech Mobile-a new service that delivers the latest Virginia Tech news and information to cellular phones or personal digital assistants (PDAs)-is now available to anyone interested in the latest news from Virginia Tech.

  4. Mahajan, Dean to lead Virginia Tech Carilion Medical Research Institute

    OpenAIRE

    Trulove, Susan

    2008-01-01

    Roop Mahajan, the James S. Tucker Professor of Engineering at Virginia Tech, has been named director of the Virginia Tech Carilion Medical Research Institute, announced Mark McNamee, Virginia Tech senior vice president and provost.

  5. Virginia Cooperative Extension and Virginia Department of Education form educational partnership

    OpenAIRE

    Burcham, Linda

    2008-01-01

    Virginia Cooperative Extension and the Virginia Department of Education have formed a partnership that will improve the educational experience of youth enrolled in family and consumer sciences programs in Virginia's public schools while also advancing the mission of both state agencies.

  6. 75 FR 24757 - Virginia Disaster #VA-00029

    Science.gov (United States)

    2010-05-05

    ... Commonwealth of Virginia (FEMA-1905-DR), dated 04/27/2010. Incident: Severe Winter Storms and Snowstorms..., Manassas Park City, Nelson, Orange, Prince William, Rappahannock, Shenandoah, Spotsylvania,...

  7. 76 FR 59765 - Virginia Disaster # VA-00036

    Science.gov (United States)

    2011-09-27

    ..., Processing and Disbursement Center, 14925 Kingsport Road, Fort Worth, TX 76155. FOR FURTHER INFORMATION...: Virginia: Charles City, Chesterfield, Colonial Heights City, Dinwiddie, Hanover, Henrico, James City,...

  8. Nuclear reactor with control rods

    International Nuclear Information System (INIS)

    The invention relates to liquid cooled nuclear reactors. In particular, it concerns reactors with mobile control rods in a straight line and guide tubes to guide these control rods through the internal upper components of the reactor vessel and in the aligned fuel assemblies of the core

  9. The Impact of Competition on Raising Mathematics Competency at Camelot Elementary School in Chesapeake, Virginia

    Science.gov (United States)

    Hayden, L. B.; Johnson, D.

    2012-12-01

    In 1995, the Virginia Department of Education approved a federal mandate for No Child Left Behind 2001 Education Act implementing the Standards of Learning (SOL) in four content areas: Mathematics, Science, English, and History and Social Sciences. These new guidelines set forth learning and achievement expectations for content areas for grades K-12 in Virginia's Public Schools. Given the SOL mandates, Virginia's elementary teachers and school leaders utilized research for specific teaching methods intended to encourage score improvements on end of year mathematics tests. In 2001, the concept of the Math Sprint Competition was introduced to Camelot Elementary School in Chesapeake Virginia, by researchers at Elizabeth City State University of Elizabeth City, North Carolina. Camelot Elementary, a K-5 school, is a Title I school nestled in a lower middle class neighborhood and houses a high number of minority students. On average, these students achieve lower test score gains than students in higher socioeconomic status district schools. Defined as a test-review based in relay format that utilizes released SOL test items, Math Sprint promotes mathematical skills outlined in Virginia SOL's and encourages competition among students that motivated them to quickly pick up on new material and retain the old material in order to out-do the others. Research identified was based on specific relationships between student competition and statewide testing results in mathematics for grades three, four, and five at Camelot Elementary. Data was compiled from results of the Math Sprint Competition and research focused on methods for motivating students encouraged by the use of a math sprint competition. Individual Pearson Product Moment Correlations were conducted to determine which variables possess strong and statistically significant relationships. Significantly, positive results came from 2005 to 2010 math sprints data from which students participated.

  10. A method and sensor for the eddy current non destructive testing of thin tubes; Procede de controle non destructif d`un tube mince par courants de Foucault

    Energy Technology Data Exchange (ETDEWEB)

    Sartre, B.; Miller, D.; Placko, D.

    1995-06-02

    In order to control the wear (cracking or thinning) of vapor generator tubes, especially in PWR reactors, due to the friction between the tubes and dampers, an eddy current control system is proposed where the transducer is run through the tubes, and measures the tube wall thickness or tube-block clearance through impedance measurements, taking into account the variation of the sensor-to-tube distance. 8 fig.

  11. Ear Tubes

    Science.gov (United States)

    ... of the ear drum or eustachian tube, Down Syndrome, cleft palate, and barotrauma (injury to the middle ear caused by a reduction of air pressure, ... specialist) may be warranted if you or your child has experienced repeated ... fluid in the middle ear, barotrauma, or have an anatomic abnormality that ...

  12. Improvement of the HANATM-4 Tubing Workability

    International Nuclear Information System (INIS)

    HANATM cladding has been developed for high burn-up fuel cladding exceeding 70,000 MWD/MTU. HIPER fuels using HANATM-6 material are currently being conducted in-reactor test in commercial nuclear reactors. HANATM-6 was produced successfully for the fuel tubing by KEPCO NF. However, the production of fuel tubing of HANATM-4 has not reached to target yield due to cracking during tube pilgering. The purpose of this study has been carried out to improve workability of HANATM-4 tubing. An improvement on the manufacturing parameters and the alloy compositions adjustments in order to improve workability HANATM-4 tubing was performed in the producing HANATM-4 cladding successfully without cracking. However, it is necessary to minor change the design of Mandrel and Die to improve the surface quality. The effects on corrosion properties and microstructure by an adjustment in manufacturing parameters and alloy compositions are currently being evaluated

  13. Study on laser welding of fuel clad tubes and end plugs made of modified 9Cr-1Mo steel for metallic fuel of Fast Breeder Reactors

    Science.gov (United States)

    Harinath, Y. V.; Gopal, K. A.; Murugan, S.; Albert, S. K.

    2013-04-01

    A procedure for Pulsed Laser Beam Welding (PLBW) has been developed for fabrication of fuel pins made of modified 9Cr-1Mo steel for metallic fuel proposed to be used in future in India's Fast Breeder Reactor (FBR) programme. Initial welding trials of the samples were carried out with different average power using Nd-YAG based PLBW process. After analyzing the welds, average power for the weld was optimized for the required depth of penetration and weld quality. Subsequently, keeping the average power constant, the effect of various other welding parameters like laser peak power, pulse frequency, pulse duration and energy per pulse on weld joint integrity were studied and a procedure that would ensure welds of acceptable quality with required depth of penetration, minimum size of fusion zone and Heat Affected Zone (HAZ) were finalized. This procedure is also found to reduce the volume fraction delta-ferrite in the fusion zone.

  14. PCR prevalence of Ranavirus in free-ranging eastern box turtles (Terrapene carolina carolina) at rehabilitation centers in three southeastern US states.

    Science.gov (United States)

    Allender, Matthew C; Abd-Eldaim, Mohamed; Schumacher, Juergen; McRuer, David; Christian, Larry S; Kennedy, Melissa

    2011-07-01

    Ranaviruses (genus Ranavirus) have been observed in disease epidemics and mass mortality events in free-ranging amphibian, turtle, and tortoise populations worldwide. Infection is highly fatal in turtles, and the potential impact on endangered populations could be devastating. Our objectives were to determine the prevalence of ranavirus DNA in blood and oral swabs, report associated clinical signs of infection, and determine spatial distribution of infected turtles. Blood and oral swabs were taken from 140 eastern box turtles (Terrapene carolina carolina) that were presented to the wildlife centers at the University of Tennessee (UT; n=39), Wildlife Center of Virginia (WCV; n=34), and North Carolina State University (NCSU; n=36), as well as a free-ranging nonrehabilitation population near Oak Ridge, Tennessee (OR; n=39) March-November 2007. Samples were evaluated for ranavirus infection using polymerase chain reaction (PCR) targeting a conserved portion of the major capsid protein. Two turtles, one from UT and one from NCSU, had evidence of ranavirus infection; sequences of PCR products were 100% homologous to Frog Virus 3. Prevalence of ranavirus DNA in blood was 3, 0, 3, and 0% for UT, WCV, NCSU, and OR, respectively. Prevalence in oral swab samples was 3, 0, and 0% for UT, WCV, and NCSU, respectively. Wildlife centers may be useful in detection of Ranavirus infection and may serve as a useful early monitoring point for regional disease outbreaks. PMID:21719848

  15. Qualitative and quantitative product analysis of the catalytic conversion of biogas in plasma-assisted flow tube reactors; Qualitative und quantitative Produktanalyse der katalytischen Konvertierung von Biogas in plasmagestuetzten Rohrstroemungsreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Kroker, Thorsten

    2011-05-31

    A cost effective method has been found to produce hydrogen and carbon monoxide from biogas by plasma enhanced flow tube. The presented method can be operated continuously (the starting material is continuously fed and the product is removed to parallel). The hydrogen and carbon monoxide are used in industry on a large scale as a synthesis gas. Synthesis gas can be e.g. refined by the Fischer-Tropsch synthesis to synthetic gasoline. In general, a synthesis gas composition of H2/CO with the ratio of 1.7 is used. Therefore a way was needed to maintain the composition of the synthesis gas, despite of the fluctuations in the CH4/CO2-composition of biogas. Due to the high flexibility of a DBE-assisted flow tube reactor, this problem could be solved (DBE = dielectric barrier discharge). [German] Es wurde ein kostenguenstiges Verfahren gefunden, das aus dem klimaunfreundlichen Biogas mit Hilfe eines plasmagestuetzten Stroemungsrohrs Wasserstoff und Kohlenmonoxid erzeugt. Das vorgestellte Verfahren kann kontinuierlich betrieben werden (das Edukt wird kontinuierlich zugefuehrt und das Produkt parallel dazu entnommen). Der Wasserstoff und das Kohlenmonoxid werden in der Industrie in grossem Massstab als Synthesegas eingesetzt. Synthesegas kann z.B. durch die Fischer-Tropsch-Synthese zu synthetischen Benzin veredelt werden. In der Regel wird dafuer eine Synthesegaszusammensetzung von H2/CO das Verhaeltnis von 1.7 eingesetzt. Deshalb musste noch ein Weg gefunden werden, um die Zusammensetzung des Synthesegases aufrechtzuerhalten, trotz Schwankungen der CH4/CO2-Zusammensetzung im Biogas. Durch die hohe Flexibilitaet eines DBE gestuetzten Stroemungsrohrreaktors konnte dieses Problem geloest werden (DBE = Dielektrische Barriere-Entladung).

  16. Underwater nuclear fuel inspection and reconstitution at Virginia Power

    International Nuclear Information System (INIS)

    Virginia Power has experienced fuel cladding defects in three of its four nuclear reactors, dating back to 1981. The first indication of fuel failure occurred at Surry Unit 1 following steam generator replacement. Subsequent examinations indicated that these failures were probably caused by debris-induced fretting. Fuel cladding defects which have occurred at the North Anna reactors are the result of both debris-induced fretting and baffle jetting. Post irradiation examinations to eliminate defective fuel assemblies have included several different techniques. A combination of wet sipping and ultrasonic testing (UT) was first used at Surry in 1983. The North Anna fuel examinations in 1984 were performed using only vacuum sipping. Each leak-detection campaign included underwater video examinations of the defective fuel assemblies and the occasional use of fibre optics and high magnification video. Because the Westinghouse fuel used by Virginia Power is not easily reconstitutable, failed fuel cannot be repaired during a refueling outage but must be replaced through a redesign of the core loading pattern using depleted fuel from the spent fuel pool. Reconstitution of a small number of once-burned fuel assemblies from Surry 1 was performed during 1985. This work occurred during a non-outage period and involved fuel assembly inversion and bottom nozzle removal. The repaired assemblies are being re-irradiated in small groups and will be used over a period of years in several fuel cycles. (author). 2 figs, 1 tab

  17. Characterization of oxides on Bruce A NGS liner tubes and steam generator tubes

    International Nuclear Information System (INIS)

    Oxide deposits on end-fitting liner tubes and steam generator tubes from the Bruce A Nuclear Generating Station (NGS) were characterized in advance of the decontamination of the heat transport system (HTS) of Bruce Unit 2. Oxide loadings, and Co-60 surface activities and specific activities were determined for the oxides on inlet and outlet end-fitting liner tubes from Bruce Unit l, Bruce Unit 2 and Bruce Unit 4. Oxides on the inner surfaces of steam generator tubes from Bruce NGS Units 1 and 2 were also characterized. The consistency in the deposit characteristics on the inlet liner tubes and steam generator tubes from Bruce A, along with the absence of magnetite on the outlet liner tubes has led to the development of a model for iron transport in the HTS of pressurized heavy water reactors (PHWRs). The activity transport/fouling mechanism involves flow-accelerated corrosion of the outlet feeder pipes, followed by deposition of iron in the steam generators, along the inlet feeder pipes, on the inlet end fittings, on the inlet fuel bundles and on the inlet region of the pressure tube. The results of loop experiments using decontamination solutions indicated that the oxide was rapidly removed from inlet liner tubes. However, removal of the Cr-rich oxide from the outlet liner tubes was less efficient, requiring the Alkaline Permangante (AP) oxidizing pre-treatment that is typically used in light water reactors (LWRs). The steam generator tubes were effectively decontaminated

  18. Ultrasonic measurement of gap between calandria tube and liquid injection shutdown system tube in PHWR

    International Nuclear Information System (INIS)

    Sag of CT or liquid injection shutdown system tubes in pressurized heavy water reactor is known to occur due to irradiation creep and growth during plant operation. When the sag of CT is big enough, the CT tube possibly comes in contact with liquid injection shutdown system tube (LIN) crossing beneath the CT, which subsequently may prevent the safe operation. It is therefore necessary to check the gap between the two tubes in order to confirm no contacts when using a proper measure periodically during the plant life. An ultrasonic gap measuring probe assembly which can be fed through viewing port installed on the calandria was developed and utilized to measure the sags of both tubes in a pressurized heavy water reactor in Korea. It was found that the centerlines of CT and LIN can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. But the measured gap data observed at the viewing port were actually not the data at the crossing point of CT and LIN. To get the actual gap between two tubes, mathematical modeling for the deflection curves of two tubes was used. The sags of CT and LIN tubes were also obtained by comparison of the present centerlines with the initial elevations at the beginning of plant operation. The gaps between two tubes in the unmeasurable regions were calculated based on the measurement data and the channel power distribution

  19. Fluidelastic instability in tube arrays

    International Nuclear Information System (INIS)

    When an array of tubes is subjected to crossflow, the tubes can experience dynamic instability, generally called fluidelastic instability. Instability initiates when the crossflow velocity exceeds a threshold value above which energy input from the flow exceeds that dissipated by system damping. Catastrophic failures of reactor and process plant equipment have been attributed to fluidelastic instability. As a result, extensive research studies have been conducted in the last 15 years with the objective of understanding the instability mechanisms and developing general design guidelines to avoid instability. Argonne National Laboratory has a continuing research program in this area which includes both mathematical model development and experimentation. This paper describes recent developments and accomplishments

  20. Electron tube

    Energy Technology Data Exchange (ETDEWEB)

    Suyama, Motohiro (Hamamatsu, JP); Fukasawa, Atsuhito (Hamamatsu, JP); Arisaka, Katsushi (Los Angeles, CA); Wang, Hanguo (North Hills, CA)

    2011-12-20

    An electron tube of the present invention includes: a vacuum vessel including a face plate portion made of synthetic silica and having a surface on which a photoelectric surface is provided, a stem portion arranged facing the photoelectric surface and made of synthetic silica, and a side tube portion having one end connected to the face plate portion and the other end connected to the stem portion and made of synthetic silica; a projection portion arranged in the vacuum vessel, extending from the stem portion toward the photoelectric surface, and made of synthetic silica; and an electron detector arranged on the projection portion, for detecting electrons from the photoelectric surface, and made of silicon.

  1. Neutron tubes

    Science.gov (United States)

    Leung, Ka-Ngo; Lou, Tak Pui; Reijonen, Jani

    2008-03-11

    A neutron tube or generator is based on a RF driven plasma ion source having a quartz or other chamber surrounded by an external RF antenna. A deuterium or mixed deuterium/tritium (or even just a tritium) plasma is generated in the chamber and D or D/T (or T) ions are extracted from the plasma. A neutron generating target is positioned so that the ion beam is incident thereon and loads the target. Incident ions cause D-D or D-T (or T-T) reactions which generate neutrons. Various embodiments differ primarily in size of the chamber and position and shape of the neutron generating target. Some neutron generators are small enough for implantation in the body. The target may be at the end of a catheter-like drift tube. The target may have a tapered or conical surface to increase target surface area.

  2. IMPACT: Virginia Potato Disease Advisory Impact

    OpenAIRE

    Herbert, D. Ames (David Ames), 1949-; Long, Theresa; Deitch, Ursula T.; Laub, Curtis A., 1955-; Rideout, Steven Lewis; Moore, David; Tucker, Lindy; Archibald, Tom

    2014-01-01

    To help potato growers more efficiently manage disease problems, Virginia Cooperative Extension initiated the Virginia Potato Disease Advisory to provide growers with weather-based disease forecast information. Advisories not only alert growers to periods of risk, they also assist growers in making management decisions, such as when protective fungicide applications are needed.

  3. VIRGINIA WILDRYE EVALUATIONS IN RIPARIAN ZONES

    Science.gov (United States)

    Virginia wildrye (Elymus virginicus L.), a perennial cool-season grass native to the northeastern USA, grows along streams, forest margins, and in other wet areas. Our previous research indicted that Virginia wildrye was not productive as a forage grass compared with introduced species such as orcha...

  4. Telelearning for Extension Agents: The Virginia Experience.

    Science.gov (United States)

    Murphy, William F., Jr.

    The creation of the Virginia Tech Teleport Facility and the installation of a nine-meter (diameter) C-Band satellite uplink antenna provided the initial impetus for the Virginia Cooperative Extension Service (VCES) to explore the use of satellite technology for information and program delivery. The $600,000 uplink became operational in September…

  5. 77 FR 47489 - Virginia Disaster #VA-00048

    Science.gov (United States)

    2012-08-08

    ... From the Federal Register Online via the Government Publishing Office SMALL BUSINESS ADMINISTRATION Virginia Disaster VA-00048 AGENCY: Small Business Administration. ACTION: Notice. SUMMARY: This is... State of Virginia (FEMA- 4072-DR), dated 07/27/2012. Incident: Severe Storms and Straight-line...

  6. Virginia Tech Bear Researchers Ensure Populations

    OpenAIRE

    Davis, Lynn

    2003-01-01

    The Cooperative Alleghany Bear Study (CABS) was initiated in 1994 as a 10-year study to ensure survival of Virginia's hunted black bear population of western Virginia. During the first six years of the study, researchers have placed radio collars on 376 of the 746 bears captured.

  7. 76 FR 37996 - West Virginia Regulatory Program

    Science.gov (United States)

    2011-06-29

    ... of approval of the West Virginia program in the January 21, 1981, Federal Register (46 FR 5915). You... Virginia is amending its Code of State Regulations (CSR) to provide for the establishment of a minimum... other things, the establishment of a minimum incremental bonding rate of $10,000 per increment at CSR...

  8. Estimation of temperature-induced reactor coolant system and steam generator tube creep rupture probability under high-pressure severe accident conditions

    International Nuclear Information System (INIS)

    A severe accident has inherently significant uncertainties due to the complex phenomena and wide range of conditions. Because of its high temperature and pressure, performing experimental validation and practical application are extremely difficult. With these difficulties, there has been few experimental researches performed and there is no plant-specific experimental data. Instead, computer codes have been developed to simulate the accident and have been used conservative assumptions and margins. This study is an effort to reduce the uncertainty in the probabilistic safety assessment and produce a realistic and physical-based failure probability. The methodology was developed and applied to the OPR1000. The creep rupture failure probabilities of reactor coolant system (RCS) components were evaluated under a station blackout severe accident with all powers lost and no recovery of steam generator auxiliary feed-water. The MELCOR 1.8.6 code was used to obtain the plant-specific pressure and temperature history of each part of the RCS and the creep rupture failure times were calculated by the rate-dependent creep rupture model with the plant-specific data. (author)

  9. Reactor control rod

    International Nuclear Information System (INIS)

    Object: To enable quick descent of a control rod body even when some relative phase deviation between upper drive means and wrapper tube is produced, while permitting a coolant to effectively flow into a protective tube irrespective of the position of the control rod body. Structure: In a control rod used for a nuclear reactor such as a fast breeder, an orifice which dispenses with a cylindrical guide tube and has a greater inner diameter than the outer diameter of the protective tube of the control rod body is provided on the inner side of a wrapper tube, thus permitting smooth operation of the control rod body and also permitting the coolant to effectively flow into the protective tube irrespective of the control rod body. (Horiuchi, T.)

  10. Morbidity and mortality of reptiles admitted to the Wildlife Center of Virginia, 1991 to 2000.

    Science.gov (United States)

    Brown, Justin D; Sleeman, Jonathan M

    2002-10-01

    Medical records from 694 reptiles admitted to the Wildlife Center of Virginia (WCV; Waynesboro, Virginia, USA) from 1991 to 2000 were reviewed to determine causes of morbidity and mortality. Eighteen species were represented but the majority of cases were four species; eastern box turtle (Terrapene carolina) (66%), eastern painted turtle (Chrysemys picta) (11%), common snapping turtle (Chelydra serpentina) (10%), and rat snake (Elaphe sp.) (6%). There was a significant increase in reptile cases during the study period both in absolute number and in proportion to the total caseload. Trauma (74%) was the most frequent cause of morbidity and mortality followed by unknown or undetermined (13%), aural abscessation (7%), infectious diseases (2%), and one nutritional disorder (0.1%). In addition, 3% of the cases were healthy animals that had been removed from the wild and consequently brought to the WCV. Causes of morbidity and mortality differed between the four most numerous species. Impact with a motor vehicle was the most frequent cause of trauma for eastern box turtles, eastern painted turtles, and common snapping turtles; however, garden-equipment-related trauma was the most frequent cause for rat snakes. Aural abscessation was only seen in eastern box turtles. Eighty percent of cases occurred between May and September and 65% occurred within the five counties closest to the WCV. The majority of morbidity and mortality was the result of human activities. The expanding human population in Virginia likely will continue to have an impact on the health of wild reptiles. PMID:12528435

  11. Development of a supermirror neutron guide tube

    International Nuclear Information System (INIS)

    A supermirror neutron guide tube with a characteristic wavelength of 1.2 A was developed and installed at the Kyoto University Reactor (KUR). The apparent critical wavelength of the present supermirrors is about 250 A. The geometrical parameters of the guide tube are : 11.7 m total length, and 10 mm wide x 74 mm high beam cross section. The supermirror neutron guide tube presents decisive advantages over the conventional nickel mirror guide tube higher transmission of neutrons, brought about through the contribution of neutrons of shorter wavelengths and of those presenting large divergent angles. The total neutron flux obtained from the KUR supermirror guide tube is about 5 x 107 n/cm2 · s, which is about 25 times what is obtainable with the conventional KUR nickel mirror neutron guide tube of the same geometrical parameters. (author)

  12. Ballooning of CANDU pressure tubes. Model assessment

    International Nuclear Information System (INIS)

    The transient creep equations used to analyze the possible ballooning and failure of Zr-2.5% Nb pressure tubes during a loss-of-coolant accident (LOCA) were developed and verified using as-received Zr-2.5% Nb pressure tube material. But in a CANDU reactor, the pressure tubes absorb deuterium and are exposed to a continuous neutron fluence. Consequently, a literature survey was done to determine how irradiation damage and deuterium might affect the creep rate and ductility of Zr-2.5% Nb pressure tubes in the temperature range from 600 to 800 degrees C. It was found that irradiation damage, dissolved deuterium and deuteride blisters could possibly affect the creep rate and ductility of ZR-2.5% Nb pressure tubes in this temperature range, but deuteride platelets are expected to have little effect. Further tests are required to determine the effect of irradiation damage and deuterium on the creep rate and ductility of pressure tubes

  13. Nuclear cluster strategy Carolinas - Ontario - Saskatchewan

    International Nuclear Information System (INIS)

    Organization of Candu Industries (OCI) is an industry association representing the interests of 170 private sector suppliers of products and services to the Canadian and offshore nuclear industries. OCI member companies, mainly in Ontario, employ over 30,000 highly specialized workers with over 12,000 working in nuclear area. OCI's objectives are to sustain the domestic nuclear program by building support among political leaders, the public and local communities, assist OCI member companies in becoming the preferred suppliers for domestic nuclear projects (competitive), assist OCI member companies in international nuclear markets - trade missions and vendor workshops. OCI is at the heart of an 'Ontario nuclear cluster'. The Carolinas have shown what can be achieved when industry, academia, S&T centers and governments collaborate with a shared vision to achieve a common goals. Ontario has the assets to become a stronger center for nuclear excellence. OCI is working to bring the pieces together. Saskatchewan has the assets to become a center of excellence in Small Modular Reactors (SMR) by licensing and constructing the first SMR in Canada.

  14. 77 FR 75448 - Welded Tube-Berkeley Including On-Site Leased Workers From Snelling, Aerotek and Express...

    Science.gov (United States)

    2012-12-20

    ... Employment and Training Administration Welded Tube--Berkeley Including On-Site Leased Workers From Snelling... Worker Adjustment Assistance on October 10, 2012, applicable to workers of Welded Tube--Berkeley... Register on October 29, 2012 (77 FR 65583). At the request of South Carolina State, the Department...

  15. Pickering NGS A: Assessment of calandria tube integrity following a sudden pressure tube failure

    International Nuclear Information System (INIS)

    The issue of calandria tube integrity following a sudden rupture of the pressure tube in Pickering NGS A reactor is addressed. Based on operating experience, only fish-mouth ruptures of the pressure tube are considered to be credible. The calandria tube response to the pressure tube break is delineated into three distinct stages, i.e. the initial transient response during the annulus filling stage, transient overpressurization and the final steady-state loading after bellows failure. The annulus response in the second stage is dominated by a waterhammer type overpressure transient with attenuation of this transient due to plastic straining of the calandria tube. The annulus pressure transients for various breaks and the sensitivity of the results to various parameters are presented. The strength margins of the calandria tube are evaluated to be relatively large. (author). 7 refs., 6 tabs., 6 figs

  16. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1981-August 1982

    International Nuclear Information System (INIS)

    The University of Virginia Reactor Facility is an integral part of the Department of Nuclear Engineering and Engineering Physics and is used to support educational programs in engineering and science at the University of Virginia and at other area colleges and universities. The University of Virginia Research Reactor (UVAR) is the highest power (two megawatts thermal power) and most utilized (total power production in 1981 and nearly 5000 megawatt-hours) research reactor in the mid-Atlantic States. In addition, a second, small (50 watt) reactor is also available for use in educational programs in the region. The University of Virginia has received support under the US Department of Energy (DOE) Reactor Sharing Program every year since 1978 to assist in meeting this objective. This report documents the major educational accomplishments under the Reactor Sharing Program for the period September 1981 through August 1982

  17. Virginia in the spotlight at the Sunbelt Agricultural Exposition

    OpenAIRE

    Greiner, Lori A.

    2007-01-01

    Virginia will be featured at the 2007 Sunbelt Agricultural Exposition to be held Oct. 16-18 in Moultrie, Ga. Virginia's exhibit, "Virginia Agriculture: Proud History, Prosperous Future," will highlight the role Virginia has played in the history of agriculture as well as showcase new directions being explored today.

  18. photomultiplier tubes

    CERN Multimedia

    photomultiplier tubes. A device to convert light into an electric signal (the name is often abbreviated to PM). Photomultipliers are used in all detectors based on scintillating material (i.e. based on large numbers of fibres which produce scintillation light at the passage of a charged particle). A photomultiplier consists of 3 main parts: firstly, a photocathode where photons are converted into electrons by the photoelectric effect; secondly, a multiplier chain consisting of a serie of dynodes which multiply the number of electron; finally, an anode, which collects the resulting current.

  19. photomultiplier tube

    CERN Multimedia

    photomultiplier tubes. A device to convert light into an electric signal (the name is often abbreviated to PM). Photomultipliers are used in all detectors based on scintillating material (i.e. based on large numbers of fibres which produce scintillation light at the passage of a charged particle). A photomultiplier consists of 3 main parts: firstly, a photocathode where photons are converted into electrons by the photoelectric effect; secondly, a multiplier chain consisting of a serie of dynodes which multiply the number of electron; finally, an anode, which collects the resulting current.

  20. Heat transfer parameters for glass-peened calandria tube in pressure tube and calandria tube contact conditions

    International Nuclear Information System (INIS)

    During a postulated event of large LOCA in CANDU reactors, the pressure tube may balloon to contact with its surrounding calandria tube to transfer heat to the moderator. To confirm the integrity of the fuel channel after the contact with a given moderator subcooling, many experiments have been performed in the last three decades by applying different pressure tube heatup rates, different pressure tube pressures and different moderator subcoolings for calandria tubes with smooth outer surface and glass-peened surface. A concept of Equivalent Moderator Subcooling (EMS) has been put forward to determine integrity of fuel channel upon pressure tube/calandria tube contact based on the existing experiment results. This concept has been presented in another work. In this work, the contact thermal conductance between pressure tube and calandria tube, critical heat flux, minimum film boiling temperature, empirical methods for nucleate boiling and film boiling heat transfer coefficient on the glass-peened calandria tube surface are discussed and estimated based on some experimental results and the EMS concept. These parameters are confirmed by simulating the existing experiments using a computer code. The estimated results may help detailed analyses on fuel channel integrity upon PT/CT contact if necessary. (author)

  1. Bearing pad to pressure tube contact simulation

    International Nuclear Information System (INIS)

    Thermal creep strain deformation is a very important pressure tube failure mechanism. During a postulated LOCA (loss of coolant accident) with failure of emergency core injection sys- tem (ECIS), the fuel cladding temperature rapidly increases and the pressure tube becomes completely dry in a few seconds after flow stagnation occurs. Subsequently, the pressure tube circumference is heated by thermal radiation except at the spots where the bearing pads are in direct contact with the pressure tube. Therefore, the localized hot spots are developed on the pressure tube's inner surface under the bearing pads. The main objective of this paper is to evaluate the local thermal-mechanical deformation of a pressure tube in a CANDU reactor and to investigate the fuel channel integrity under localized contact between bearing pad and pressure tube. Furthermore, the mechanistic models are validated against the experimental works per- formed at WRL (Whiteshell research laboratory). Calculations are performed using the finite element method in which the heat, thermal mechanical and creep strain equations are solved, simultaneously. According to the experimental set up, the heat conduction from bearing pads to the inner surface of the pressure tube with appropriate convective and radiation boundary conditions has been simulated. Furthermore, the thermal creep strain deformation has been obtained for when the pressure tube is still under operational condition. It is observed that the pressure tube thermal strain will occur if sufficient high temperature is reached however, depending on the severity of flow degradation in the fuel channel, these localized hot spots could represent a potential creep strain failure of the pressure tube. Whether the pressure tube would fail at these hot spots before contacting the calandria tube depends on the localized temperature and experienced pressure transients. Sensitivity analysis is performed in order to evaluate the contact conductance, the

  2. Manufacturing and testing the HTGR refueling tube

    International Nuclear Information System (INIS)

    The paper describes the manufacturing technique for a refueling tube of a high-temperature gas-cooled nuclear reactor (HTGR). Four refueling tube sections were made: two sections from GSP-50 material and two sections from carbon-carbon (C-C) composite materials. Radiation tests were carried out in the reactor BOR-60. Experimental results show that the strength characteristics and thermophysical properties of graphitized carbon materials, from which the sections have been manufactured, are higher by a factor of 2.5-3.5 as compared with the HTGR refueling tube requirements. The dimensional changes of GSP-50 and C-C composite materials at temperatures between 300 and 600 deg C up to the neutron fluence of 1·1021 n/cm2 are comparable and meet the specifications for HTGR refueling tube

  3. A Foray into Library Digital Publishing: The British Virginia Project at Virginia Commonwealth University

    OpenAIRE

    Farley,Kevin

    2014-01-01

    The British Virginia project involves a collaboration between Virginia Commonwealth University (VCU) Libraries and faculty members in the departments of English and History at VCU, with the project led by Dr. Joshua Eckhardt (English). As of April 25, 2013, the project has published its first title: an online edition of a sermon preached to the Virginia Company by William Symonds. To ensure the success of this project, a number of details required careful planning, including library outreach,...

  4. Nuclear reactor with a reactor core composed of fuel elements

    International Nuclear Information System (INIS)

    A tube surrounding a fuel element projects above the liquid level. The tube is situated in a pot, whose upper edge lies between the top of the reactor core and the liquid level. A greater pressure is therefore produced, which ensures a reduction of the steam bubble proportion in the cooling liquid at the other fuel elements. (orig./HP)

  5. Health and safety impact of steam generator tube degradation

    International Nuclear Information System (INIS)

    In this paper the author addresses the problems inherent in evaluating the safety of steam generators with respect to tube rupture as part of a probabilistic safety analysis (PSA) of a reactor plant. He reviews the history of PSA as applied to reactors, and then looks at tube rupture histories as a start toward establishing event frequencies. He considers tube ruptures from the aspect of being an initiating event to being a conditional event to some other event, and then the question of performance of the steam generator in the face of a severe accident in the reactor

  6. Health and safety impact of steam generator tube degradation

    Energy Technology Data Exchange (ETDEWEB)

    Marston T. [PLG, Inc., Newport Beach, CA (United States)

    1997-02-01

    In this paper the author addresses the problems inherent in evaluating the safety of steam generators with respect to tube rupture as part of a probabilistic safety analysis (PSA) of a reactor plant. He reviews the history of PSA as applied to reactors, and then looks at tube rupture histories as a start toward establishing event frequencies. He considers tube ruptures from the aspect of being an initiating event to being a conditional event to some other event, and then the question of performance of the steam generator in the face of a severe accident in the reactor.

  7. North Carolina School Nurse Leadership Institute

    Science.gov (United States)

    Guttu, Martha

    2007-01-01

    Recognizing that school nurse leaders are essential to the development of school nurses, the North Carolina School Nurse Leadership Institute was developed to enable school nurse leaders to update and advance their leadership skills. The Institute was a collaborative endeavor between the North Carolina Department of Health and Human Services,…

  8. Financial Flexibility in North Carolina Schools.

    Science.gov (United States)

    Suarez, Tanya M.; Polen, Deborah A.

    This paper explores educational financial flexibility with a focus on the specific issues surrounding local flexibility in North Carolina school districts. Strategies that states have used to increase local financial flexibility include waivers, reduction of budget categories, block grants, and school-based budgeting. The North Carolina system of…

  9. A Guide for Virginia's Forest Landowners

    OpenAIRE

    Munsell, John F; Gagnon, Jennifer L.

    2011-01-01

    Provides Virginia's forest landowners with strategies for making profitable investments through sustainable forest management. It describes the seven-criteria Montreal Process and lists sources of information and assistance.

  10. Physics launches first Virginia Tech signature experience

    OpenAIRE

    Doss, Catherine

    2010-01-01

    When Professor Nahum Arav joined the Department of Physics in the College of Science in January 2008, he says he had a dream: to introduce Virginia Tech students to the beauty and wonders of the universe.

  11. Virginia ESI: MGT (Management Area Polygons)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains boundaries for management areas, national parks, state and local parks, and wildlife refuges in Virginia. Vector polygons in this data set...

  12. Virginia Woolfi tunnid kummitavad edasi / Andres Laasik

    Index Scriptorium Estoniae

    Laasik, Andres, 1960-2016

    2003-01-01

    Virginia Woolfi teosest "Mrs. Dalloway" ajendatud Michael Cunninghami romaanil "Tunnid" põhinev mängufilm "Tunnid" ("The Hours") : režissöör Stephen Daldry : kesksetes rollides Meryl Streep, Nicole Kidman, Julianne Moore : Suurbritannia 2002

  13. Recently Deceased: The First Amendment in Virginia.

    Science.gov (United States)

    Meyers, Terry L.

    2002-01-01

    Describes how a Virginia law forbidding state employees to access sexually explicit material on state-owned computers without written permission from their agency's head has "plagued" one professor, requiring numerous incidents of permission-seeking regarding Victorian poetry. (EV)

  14. 75 FR 9006 - Virginia Disaster #VA-00028

    Science.gov (United States)

    2010-02-26

    ... Commonwealth of Virginia (FEMA-1874-DR), dated 02/16/2010. Incident: Severe Winter Storm and Snowstorm... Park City, Montgomery, Nelson, Norton City, Orange, Page, Prince William, Rockbridge, Russell,...

  15. Virginia Tech partnership wins NASPA award

    OpenAIRE

    Harris, Sally L.

    2005-01-01

    A program fostered and nurtured by Landrum Cross, Virginia Tech's vice president for Student Affairs, has won the Global Partnership Program award in NASPA International Education Knowledge Community's 3rd Annual Best Practices in International Education and Learning Awards.

  16. ORTHOIMAGERY, CITY OF POQUOSON, VIRGINIA, USA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — These files contain Digital Orthophoto files for the State of Virginia developed from imagery acquired in spring 2007. In the spring of 2006, the Commonwealth of...

  17. Growing Peaches and Nectarines in Virginia

    OpenAIRE

    Marini, Richard P. (Richard Paul), 1952-

    2009-01-01

    Discusses growing peaches and nectarines in Virginia, and offers advice on where best to locate the orchard, selection of varieties, care of trees, soil preparation, weed control, fertilization, fruit thinning, pest control, harvest, and storage.

  18. 2012 FEMA Lidar: Southern Virginia Counties

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Dewberry collected LiDAR for ~3,341 square miles in various Virginia Counties, a part of Worcester County, and Hoopers Island. The acquisition was performed by...

  19. 2011 FEMA Lidar: Southern Virginia Cities

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Dewberry collected LiDAR for ~3,341 square miles in various Virginia Counties, a part of Worcester County, and Hooper's Island. The acquisition was performed by...

  20. H09738: NOS Hydrographic Survey , Offshore Atlantic Ocean, Virginia and North Carolina, 1978-03-21

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The National Oceanic and Atmospheric Administration (NOAA) has the statutory mandate to collect hydrographic data in support of nautical chart compilation for safe...