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Sample records for carolinas virginia tube reactor

  1. Pressure tube reactor

    International Nuclear Information System (INIS)

    Matsumoto, Tomoyuki; Fujino, Michihira.

    1980-01-01

    Purpose: To equalize heavy water flow distribution by providing a nozzle for externally injecting heavy water from a vibration preventive plate to the upper portion to feed the heavy water in a pressure tube reactor and swallowing up heavy water in a calandria tank to supply the heavy water to the reactor core above the vibration preventive plate. Constitution: A moderator injection nozzle is mounted on the inner wall of a calandria tank. Heavy water is externally injected above the vibration preventive plate, and heavy water in the calandria tank is swallowed up to supply the heavy water to the core reactor above the vibration preventive plate. Therefore, the heavy water flow distribution can be equalized over the entire reactor core, and the distribution of neutron absorber dissolved in the heavy water is equalized. (Yoshihara, H.)

  2. Ecology of Virginia big-eared bats in North Carolina and Tennessee.

    Science.gov (United States)

    2016-08-24

    The researchers conducted a study of the springtime ecology of an isolated North Carolina-Tennessee population of the Virginia big-eared bat (Corynorhinus townsendii virginianus), a federally endangered species. With limited data on the whereabouts o...

  3. 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.

  4. A new species of Perlesta (Plecoptera: Perlidae) from North Carolina with additional records for North Carolina and Virginia

    Science.gov (United States)

    Kondratieff, B.C.; Zuellig, R.E.; Lenat, D.R.

    2011-01-01

    Twenty-eight species of Nearctic Perlesta are currently recognized (Stark 1989, 2004; Kondratieff et al. 2006, 2008; Grubbs and DeWalt 2008, Grubbs and DeWalt 2011, Kondratieff and Myers 2011). Interestingly, but needing confirmation, Perlesta has been recently recorded from Central America (Gutiérrez-Fonseca and Springer 2011). Continued collecting and study of Perlesta from North Carolina by the authors revealed one additional undescribed species. Ten species of Perlesta currently have been recorded from North Carolina (Stark 1989, 2004, Kondratieff et al. 2006, 2008, Grubbs and DeWalt 2008). Additionally, new Perlesta species records are given for Virginia. The terminology used in the description of the male adult follows Stark (1989, 2004).

  5. Improving the calandria tubes for CANDU reactors

    International Nuclear Information System (INIS)

    Coleman, C.E.; Fong, R.W.L.; Doubt, G.L.

    1997-01-01

    CANDU calandria tubes are made from annealed Zircaloy-2 sheet formed into a cylinder and welded along its length to make the tube. The current calandria tubes have given exemplary service for many years. With more stringent regulations and the need to accommodate warm cooling water in tropical countries, we started a development program to increase the margins for failure during postulated accidents. These improvements involve increasing the tube strength and optimising the heat-transfer from an excessively hot fuel channel to the cool moderator. If the postulated accident involves a pressure tube break, it would be desirable if the calandria tube withstood the full pressure of the heat-transport system. The weakest link in current calandria tubes is the weld. Thickening the weld can increase the strength by 20% while seamless tubes can be 45% stronger than current tubes. The latter tubes can hold full system pressure for many hours without failure. If during the postulated accident the fuel and pressure tube become excessively hot but do not touch the calandria tube, the radiant heat loss must be maximised. Current calandria tubes have an absorptivity (emissivity) of about 0.2. To protect the fuel and the fuel channel we have devised a finish to the inside surface of the calandria tube that increases the emissivity to 0.7. If during the postulated accident the hot pressure tube touches the cool calandria tube, the contact conductance and the critical heat flux must be optimised to ensure nucleate boiling of the moderator at the outside surface of the calandria tube and therefore efficient exploitation of the moderator as a heat sink. In laboratory tests small ridges on the inside surface and roughening of the outside surface have been shown to increase the margins against failure and increase the possible moderator temperatures thus providing the opportunity to decrease the cost of the moderator heat-exchange system and remove restrictions on reactor operation in

  6. An analysis, using the CLAPTRAP code, of the pressure transients developed in the Carolinas Virginia Tube Reactor during containment performance tests

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1982-11-01

    To check containment performance of the CVTR, steam was injected above the operating floor through a 10 foot pipe cap containing the 1 inch diameter holes, at a steady rate of 102.8 lb/sec for a period of 166 seconds. This steam had an enthalpy of 1195 Btu/lb and was therefore not entirely typical of the much wetter material which would be rejected for the greater part of a true breached circuit accident. Pressure transients measured experimentally within the containment were compared with results calculated by the American code CONTEMPT and these results in turn have allowed the Winfrith code CLAPTRAP to be tested for consistency and to establish that the use of this code would have led to similar conclusions about the heat transfer coefficients at the heat absorbent surfaces. (U.K.)

  7. Calandria cooling structure in pressure tube reactor

    International Nuclear Information System (INIS)

    Hyugaji, Takenori; Sasada, Yasuhiro.

    1976-01-01

    Purpose: To contrive the structure of a heavy water distributing device in a pressure tube reactor thereby to reduce the variation in the cooling function thereof due to the welding deformation and installation error. Constitution: A heating water distributing plate is provided at the lower part of the upper tubular plate of a calandria tank to form a heavy water distributing chamber between both plates and a plurality of calandria tubes. Heavy water which has flowed in the upper part of the heavy water distributing plate from the heavy water inlet nozzle flows down through gaps formed around the calandria tubes, whereby the cooling of the calandria tank and the calandria tubes is carried out. In the above described calandria cooling structure, a heavy water distributing plate support is provided to secure the heavy water distributing plate and torus-shaped heavy water distributing rings are fixed to holes formed in the heavy water distributing plate penetrating through the calandria tubes thereby to form torus-shaped heavy water outlet ports each having a space. (Seki, T.)

  8. Performance of pressure tubes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

    2016-06-15

    The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

  9. Ground Water Atlas of the United States: Segment 11, Delaware, Maryland, New Jersey, North Carolina, Pennsylvania, Virginia, West Virginia

    Science.gov (United States)

    Trapp, Henry; Horn, Marilee A.

    1997-01-01

    Segment 11 consists of the States of Delaware, Maryland, New Jersey, North Carolina, West Virginia, and the Commonwealths of Pennsylvania and Virginia. All but West Virginia border on the Atlantic Ocean or tidewater. Pennsylvania also borders on Lake Erie. Small parts of northwestern and north-central Pennsylvania drain to Lake Erie and Lake Ontario; the rest of the segment drains either to the Atlantic Ocean or the Gulf of Mexico. Major rivers include the Hudson, the Delaware, the Susquehanna, the Potomac, the Rappahannock, the James, the Chowan, the Neuse, the Tar, the Cape Fear, and the Yadkin-Peedee, all of which drain into the Atlantic Ocean, and the Ohio and its tributaries, which drain to the Gulf of Mexico. Although rivers are important sources of water supply for many cities, such as Trenton, N.J.; Philadelphia and Pittsburgh, Pa.; Baltimore, Md.; Washington, D.C.; Richmond, Va.; and Raleigh, N.C., one-fourth of the population, particularly the people who live on the Coastal Plain, depends on ground water for supply. Such cities as Camden, N.J.; Dover, Del.; Salisbury and Annapolis, Md.; Parkersburg and Weirton, W.Va.; Norfolk, Va.; and New Bern and Kinston, N.C., use ground water as a source of public supply. All the water in Segment 11 originates as precipitation. Average annual precipitation ranges from less than 36 inches in parts of Pennsylvania, Maryland, Virginia, and West Virginia to more than 80 inches in parts of southwestern North Carolina (fig. 1). In general, precipitation is greatest in mountainous areas (because water tends to condense from moisture-laden air masses as the air passes over the higher altitudes) and near the coast, where water vapor that has been evaporated from the ocean is picked up by onshore winds and falls as precipitation when it reaches the shoreline. Some of the precipitation returns to the atmosphere by evapotranspiration (evaporation plus transpiration by plants), but much of it either flows overland into streams as

  10. NOAA submerged aquatic vegetation (SAV) habitat mapping orthoimagery, collection subset 1 of 2, coastal North Carolina and SE Virginia, 2007-2008 (NODC Accession 0086096)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Orthophotography was flown in coastal regions of North Carolina and southeastern Virginia in an effort to establish long term mapping and monitoring of submerged...

  11. NOAA submerged aquatic vegetation (SAV) habitat mapping orthoimagery, collection subset 2 of 2, coastal North Carolina and SE Virginia, 2007-2008 (NODC Accession 0086104)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Orthophotography was flown in coastal regions of North Carolina and southeastern Virginia in an effort to establish long term mapping and monitoring of submerged...

  12. Post-Hurricane Irene coastal oblique aerial photographs collected from Ocracoke Inlet, North Carolina, to Virginia Beach, Virginia, August 30-31, 2011

    Science.gov (United States)

    Morgan, Karen L. M.; Krohn, M. Dennis

    2016-02-17

    The U.S. Geological Survey (USGS), as part of the National Assessment of Coastal Change Hazards project, conducts baseline and storm-response photography missions to document and understand the changes in vulnerability of the Nation's coasts to extreme storms (Morgan, 2009). On August 30-31, 2011, the USGS conducted an oblique aerial photographic survey from Ocracoke Inlet, North Carolina, to Virginia Beach, Virginia, aboard a Piper Navajo Chieftain (aircraft) at an altitude of 500 feet (ft) and approximately 1,200 ft offshore. This mission was flown to collect post-Hurricane Irene data for assessing incremental changes in the beach and nearshore area since the last survey, flown in May 2008, and the data can be used in the assessment of future coastal change.

  13. Winter 2016, Part A—Coastal oblique aerial photographs collected from the South Carolina/North Carolina border to Assateague Island, Virginia, February 18–19, 2016

    Science.gov (United States)

    Morgan, Karen L. M.

    2017-02-28

    The U.S. Geological Survey (USGS), as part of the National Assessment of Coastal Change Hazards project, conducts baseline and storm-response photography missions to document and understand the changes in the vulnerability of the Nation's coasts to extreme storms. On February 18–19, 2016, the USGS conducted an oblique aerial photographic survey from the South Carolina/North Carolina border to Assateague Island, Virginia, aboard a Cessna 182 (aircraft) 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 in the beach and nearshore area and can be used to assess future coastal change.The photographs in this report document the state of the barrier islands and other coastal features at the time of the survey.

  14. Eddy current measurement of remote tube positions in CANDU reactors

    International Nuclear Information System (INIS)

    Craig, S.T.; Krause, T.W.; Luloff, B.V.; Schankula, J.J.

    2004-01-01

    Regular NDE inspections of CANDU reactors are made by inserting probes into one of 380 pressure tubes that traverse the core. Each pressure tube is surrounded by a gas annulus and contained in a calandria tube. Separation between the pressure tube and calandria tubes is maintained by four spacers. Moderator water surrounds the calandria tubes. Auxiliary tubular assemblies within the moderator run perpendicular to the pressure tube for the injection of neutron poison and for mounting flux detectors. Laboratory tests have demonstrated the use of remote field eddy currents to measure the distance between the pressure tube and auxiliary tubes. Implementation uses coils from two appropriately separated probes in existing inspection heads. These coils are the transmit coil used for sensing the spacers and the receive coil from a probe used to measure the gap between the pressure tube and its surrounding calandria tube. The axis of the transmit coil is aligned with the axis of the pressure tube. The receive coil axis is perpendicular to the transmit coil, and located near the inner diameter of the pressure tube. Although the coil spacing and orientation are not ideal, laboratory tests have demonstrated repeatable measurements under conditions of varying liftoff, pressure tube wall thickness and diameter, and gap between pressure tube and calandria tube. The experimental conditions, test cases, and results are presented. (author)

  15. 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

  16. Strategic conservation planning for the Eastern North Carolina/Southeastern Virginia Strategic Habitat Conservation Team

    Science.gov (United States)

    Alexander-Vaughn, Louise B.; Collazo, Jaime A.; Drew, C. Ashton

    2014-01-01

    The Eastern North Carolina/Southeastern Virginia Strategic Habitat Conservation Team (ENCSEVA) is a partnership among local federal agencies and programs with a mission to apply Strategic Habitat Conservation to accomplish priority landscape-level conservation within its geographic region. ENCSEVA seeks to further landscape-scale conservation through collaboration with local partners. To accomplish this mission, ENCSEVA is developing a comprehensive Strategic Habitat Conservation Plan (Plan) to provide guidance for its members, partners, and collaborators by establishing mutual conservation goals, objectives, strategies, and metrics to gauge the success of conservation efforts. Identifying common goals allows the ENCSEVA team to develop strategies that leverage joint resources and are more likely to achieve desired impacts across the landscape. The Plan will also provide an approach for ENCSEVA to meet applied research needs (identify knowledge gaps), foster adaptive management principles, identify conservation priorities, prioritize threats (including potential impacts of climate change), and identify the required capacity to implement strategies to create more resilient landscapes. ENCSEVA seeks to support the overarching goals of the South Atlantic Landscape Conservation Cooperative (SALCC) and to provide scientific and technical support for conservation at landscape scales as well as inform the management of natural resources in response to shifts in climate, habitat fragmentation and loss, and other landscape-level challenges (South Atlantic LCC 2012). The ENCSEVA ecoregion encompasses the northern third of the SALCC geography and offers a unique opportunity to apply landscape conservation at multiple scales through the guidance of local conservation and natural resource management efforts and by reporting metrics that reflect the effectiveness of those efforts (Figure 1). The Environmental Decision Analysis Team, housed within the North Carolina Cooperative

  17. 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.

  18. Northward extension of Carolina slate belt stratigraphy and structure, South-Central Virginia: Results from geologic mapping

    Science.gov (United States)

    Hackley, P.C.; Peper, J.D.; Burton, W.C.; Horton, J. Wright

    2007-01-01

    Geologic mapping in south-central Virginia demonstrates that the stratigraphy and structure of the Carolina slate belt extend northward across a steep thermal gradient into upper amphibolite-facies correlative gneiss and schist. The Neoproterozoic greenschist-facies Hyco, Aaron, and Virgilina Formations were traced northward from their type localities near Virgilina, Virginia, along a simple, upright, northeast-trending isoclinal syncline. This syncline is called the Dryburg syncline and is a northern extension of the more complex Virgilina synclinorium. Progressively higher-grade equivalents of the Hyco and Aaron Formations were mapped northward along the axial trace of the refolded and westwardly-overturned Dryburg syncline through the Keysville and Green Bay 7.5-minute quadrangles, and across the northern end of the Carolina slate belt as interpreted on previous geologic maps. Hyco rocks, including felsic metatuff, metawacke, and amphibolite, become gneisses upgrade with areas of local anatexis and the segregation of granitic melt into leucosomes with biotite selvages. Phyllite of the Aaron Formation becomes garnet-bearing mica schist. Aaron Formation rocks disconformably overlie the primarily felsic volcanic and volcaniclastic rocks of the Hyco Formation as evidenced by repeated truncation of internal contacts within the Hyco on both limbs of the Dryburg syncline at the Aaron-Hyco contact. East-northeast-trending isograds, defined successively by the first appearance of garnet, then kyanite ?? staurolite in sufficiently aluminous rocks, are superposed on the stratigraphic units and synclinal structure at moderate to high angles to strike. The textural distinction between gneisses and identifiable sedimentary structures occurs near the kyanite ?? staurolite-in isograd. Development of the steep thermal gradient and regional penetrative fabric is interpreted to result from emplacement of the Goochland terrane adjacent to the northern end of the slate belt during

  19. Visual beam tube inspection at the TRIGA reactor Vienna

    International Nuclear Information System (INIS)

    Boeck, H.; Musilek, A.; Villa, M.

    2006-01-01

    Of the four TRIGA beam tubes two have been visually inspected in 1985. Prior to the inspection the reactor was shut down for 3 weeks. The fuel elements around the beam tubes were removed. Stainless steel dummy elements were inserted in the fuel positions to shield the core radiation. The active part of the Fast Rabbit Tube was removed into the beam tube loading device and transferred to an interim storage: Front dose rate was ∼ 50 mSv/h. Generally the beam tube was very clean, after the last inspection about 30 years ago. A1 cm cut was observed at the beam tube front end. A rigid endoscope was used to check the beam tube's inner surface using a 90 degree deflection objective and photo- and video equipment. The direct dose rate in front of the beam tube was about 30 mSv/h. The beam tube was vacuum cleaned. A corroded shielding tank containing boric acid has leaked. A wooden collimator partially disintegrating due to extreme temperature was removed from beam tube D. Documentation of the inspection for visible defects is produced for later comparison

  20. University of Virginia open-quotes virtualclose quotes reactor facility tours

    International Nuclear Information System (INIS)

    Krause, D.R.; Mulder, R.U.

    1995-01-01

    An electronic information and tour book has been constructed for the University of Virginia reactor (UVAR) facility. Utilizing the global Internet, the document resides on the University of Virginia World Wide Web (WWW or W) server within the UVAR Homepage at http://www.virginia. edu/∼reactor/. It is quickly accessible wherever an Internet connection exists. The UVAR Homepage files are accessed with the hypertext transfer protocol (http) prefix. The files are written in hypertext markup language (HTML), a very simple method of preparing ASCII text for W3 presentation. The HTML allows use of various hierarchies of headers, indentation, fonts, and the linking of words and/or pictures to other addresses-uniform resource locators. The linking of texts, pictures, sounds, and server addresses is known as hypermedia

  1. Fabrication characteristics of zircaloy tubes for nuclear reactors

    International Nuclear Information System (INIS)

    Haydt, H.M.

    1980-11-01

    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) [pt

  2. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1985-01-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in PWRs, formed the basis of study for the last year of the project. Four tasks are addressed in this study of the detection of steam tube leaks. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above. (4) Assessing the need for diagnostic data processing and display. Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation. (orig./HP)

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

    International Nuclear Information System (INIS)

    Pease, L.

    1977-08-01

    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)

  4. Sediment distribution and hydrologic conditions of the Potomac aquifer in Virginia and parts of Maryland and North Carolina

    Science.gov (United States)

    McFarland, Randolph E.

    2013-01-01

    Sediments of the heavily used Potomac aquifer broadly contrast across major structural features of the Atlantic Coastal Plain Physiographic Province in eastern Virginia and adjacent parts of Maryland and North Carolina. Thicknesses and relative dominance of the highly interbedded fluvial sediments vary regionally. Vertical intervals in boreholes of coarse-grained sediment commonly targeted for completion of water-supply wells are thickest and most widespread across the central and southern parts of the Virginia Coastal Plain. Designated as the Norfolk arch depositional subarea, the entire sediment thickness here functions hydraulically as a single interconnected aquifer. By contrast, coarse-grained sediment intervals are thinner and less widespread across the northern part of the Virginia Coastal Plain and into southern Maryland, designated as the Salisbury embayment depositional subarea. Fine-grained intervals that are generally avoided for completion of water-supply wells are increasingly thick and widespread northward. Fine-grained intervals collectively as thick as several hundred feet comprise two continuous confining units that hydraulically separate three vertically spaced subaquifers. The subaquifers are continuous northward but merge southward into the single undivided Potomac aquifer. Lastly, far southeastern Virginia and northeastern North Carolina are designated as the Albemarle embayment depositional subarea, where both coarse- and fine-grained intervals are of only moderate thickness. The entire sediment thickness functions hydraulically as a single interconnected aquifer. A substantial hydrologic separation from overlying aquifers is imposed by the upper Cenomanian confining unit. Potomac aquifer sediments were deposited by a fluvial depositional complex spanning the Virginia Coastal Plain approximately 100 to 145 million years ago. Westward, persistently uplifted granite and gneiss source rocks sustained a supply of coarse-grained sand and gravel

  5. Low in reactor creep Zr-base alloy tubes

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Holt, R.A.

    1984-01-01

    This invention relates to zirconium alloy tubes especially for use in nuclear power reactors. More particularly it relates to quaternary 3.5 percent Sn, 1 percent Mo, 1 percent Nb, balance Zr alloy tubes which have been extruded, cold worked and heat treated to lower their dislocation density. In one embodiment the alloys are cold worked less than 5 percent and stress relieved to produce a low dislocation density and in another embodiment the alloys are cold worked up to about 50 percent and annealed to produce a very low dislocation density and also small equiaxed β grains

  6. Core construction in a pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Ueda, Makoto; Aoki, Katsutada.

    1975-01-01

    Object: To replace a centrally positioned fuel assembly of a fuel assembly unit with a reactor controlling machinery to decrease a distance between the fuel assemblies thereby saving use of heavy water and enhancing economy. Structure: A centrally positioned fuel assembly of a fuel assembly unit, which is composed of a plurality of fuel assemblies orderly arranged in lattice fashion, is replaced with a reactor controlling members such as control rods, poison tubes and the like to provide an arrangement of lattice-free type fuel assembly, thus reducing the pitch as small as possible. (Kamimura, M.)

  7. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1984-11-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It identifies physical parameters, establishes instrumentation performance goals, and specifies sensor types and locations. It presents a simple algorithm that yields the leak rate as a function of known or measurable quantities. Leak rates of less than one-tenth gram per second should be detectable with existing instrumentation

  8. 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...

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

    International Nuclear Information System (INIS)

    Tatone, O.S.; Tapping, R.L.

    1988-12-01

    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

  10. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

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

    International Nuclear Information System (INIS)

    Mayo, C.W.

    1992-01-01

    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 UO 2 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

  12. Automated Determination of Oxygen-Dependent Enzyme Kinetics in a Tube-in-Tube Flow Reactor

    DEFF Research Database (Denmark)

    Ringborg, Rolf Hoffmeyer; Pedersen, Asbjørn Toftgaard; Woodley, John

    2017-01-01

    and limited oxygen supply. Here, we present a novel method for the collection of such kinetic data using a pressurized tube-in-tube reactor, operated in the low-dispersed flow regime to generate time-series data, with minimal material consumption. Experimental development and validation of the instrument......Enzyme-mediated oxidation is of particular interest to synthetic organic chemists. However, the implementation of such systems demands knowledge of enzyme kinetics. Conventionally collecting kinetic data for biocatalytic oxidations is fraught with difficulties such as low oxygen solubility in water...... revealed not only the high degree of accuracy of the kinetic data obtained, but also the necessity of making measurements in this way to enable the accurate evaluation of high KMO enzyme systems. For the first time, this paves the way to integrate kinetic data into the protein engineering cycle....

  13. Eddy current proximity measurement of perpendicular tubes from within pressure tubes in CANDU nuclear reactors

    Science.gov (United States)

    Bennett, P. F. D.; Underhill, P. R.; Morelli, J.; Krause, T. W.

    2018-04-01

    Fuel channels in CANDU® (CANada Deuterium Uranium) nuclear reactors consist of two non-concentric tubes; an inner pressure tube (PT) and a larger diameter calandria tube (CT). Up to 400 horizontally mounted fuel channels are contained within a calandria vessel, which also holds the heavy water moderator. Certain fuel channels pass perpendicularly over horizontally oriented tubes (nozzles) that are part of the reactor's liquid injection shutdown system (LISS). Due to sag, these fuel channels are at risk of coming into contact with the LISS nozzles. In the event of contact between the LISS nozzle and CT, flow-induced vibrations from within the moderator could lead to fretting and deformation of the CT. LISS nozzle proximity to CTs is currently measured optically from within the calandria vessel, but from outside the fuel channels. Measurement by an independent means would provide confidence in optical results and supplement cases where optical observations are not possible. Separation of PT and CT, known as gap, is monitored from within the PT using a transmit-receive eddy current probe. Investigation of the eddy current based gap probe as a tool to also measure proximity of LISS nozzles was carried out experimentally in this work. Eddy current response as a function of LISS-PT proximity was recorded. When PT-CT gap, PT wall thickness, PT resistivity and probe lift-off variations were not present this dependence could be used to determine the LISS-PT proximity. This method has the potential to provide LISS-CT proximity using existing gap measurement data. Obtaining LISS nozzle proximity at multiple inspection intervals could be used to provide an estimate of the time to LISS-CT contact, and thereby provide a means of optimizing maintenance schedules.

  14. Detailed geochemical study of the Dan River-Danville Triassic Basin, North Carolina and Virginia. National Uranium Resource Evaluation Program

    International Nuclear Information System (INIS)

    Thayer, P.A.; Cook, J.R.

    1982-08-01

    This abbreviated data report presents results of surface geochemical reconnaissance in the Dan River-Danville Triassic Basin of north-central North Carolina and south-central Virginia. Unweathered rock samples were collected at 380 sites within the basin at a nominal sampling density of one site per square mile. Field measurements and observations are reported for each site; analytical data and field measurements are presented in tables and maps. A detailed four-channel spectrometric survey was conducted, and the results are presented as a series of symbol plot maps for eU, eTh, and eU/eTh. Data from rock sample sites (on microfiche in pocket) include rock type and color and elemental analyses for U, Th, Hf, Al, Ce, Dy, Eu, Fe, La, Lu, Mn, Na, Sc, Sm, Ti, V, and Yb. Elemental uranium in 362 sedimentary rock samples from the Dan River-Danville Basin ranges from a low of 0.1 to a maximum of 13.3 parts per million (ppM). The log mean uranium concentration for these same samples is 0.37 ppM, and the log standard deviation is 0.24 ppM. Elemental uranium in 10 diabase dike samples from within the basin is in the range 0.1 to 0.7 ppM. The log mean uranium concentration for diabase samples is -.65 ppM, and the log standard deviation is 0.27. This report is issued in draft form, without detailed technical and copy editing. This was done to make the report available to the public before the end of the NURE program

  15. Flow chemistry: intelligent processing of gas-liquid transformations using a tube-in-tube reactor.

    Science.gov (United States)

    Brzozowski, Martin; O'Brien, Matthew; Ley, Steven V; Polyzos, Anastasios

    2015-02-17

    reactive gas in a given reaction mixture. We have developed a tube-in-tube reactor device consisting of a pair of concentric capillaries in which pressurized gas permeates through an inner Teflon AF-2400 tube and reacts with dissolved substrate within a liquid phase that flows within a second gas impermeable tube. This Account examines our efforts toward the development of a simple, unified methodology for the processing of gaseous reagents in flow by way of development of a tube-in-tube reactor device and applications to key C-C, C-N, and C-O bond forming and hydrogenation reactions. We further describe the application to multistep reactions using solid-supported reagents and extend the technology to processes utilizing multiple gas reagents. A key feature of our work is the development of computer-aided imaging techniques to allow automated in-line monitoring of gas concentration and stoichiometry in real time. We anticipate that this Account will illustrate the convenience and benefits of membrane tube-in-tube reactor technology to improve and concomitantly broaden the scope of gas/liquid/solid reactions in organic synthesis.

  16. Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II

    International Nuclear Information System (INIS)

    2010-01-01

    The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

  17. Tube-in-tube reactor as a useful tool for homo- and heterogeneous olefin metathesis under continuous flow mode.

    Science.gov (United States)

    Skowerski, Krzysztof; Czarnocki, Stefan J; Knapkiewicz, Paweł

    2014-02-01

    A tube-in-tube reactor was successfully applied in homo- and heterogeneous olefin metathesis reactions under continuous flow mode. It was shown that the efficient removal of ethylene facilitated by connection of the reactor with a vacuum pump significantly improves the outcome of metathesis reactions. The beneficial aspects of this approach are most apparent in reactions performed at low concentration, such as macrocyclization reactions. The established system allows achievement of both improved yield and selectivity, and is ideal for industrial applications. Copyright © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  18. Thin-walled large-diameter zirconium alloy tubes in CANDU reactors

    International Nuclear Information System (INIS)

    Price, E.G.; Richinson, P.J.

    1978-08-01

    The requirements of the thin-walled large-diameter Zircaloy-2 tubing used in CANDU reactors are reviewed. Strength, residual stress patterns, texture and prior deformation contribute to the stability of these tubes. The extent to which the present manufacturing route meets these requirements is discussed. (author)

  19. Lack of healthy food options on children's menus of restaurants in the health-disparate Dan River region of Virginia and North Carolina, 2013.

    Science.gov (United States)

    Hill, Jennie L; Olive, Nicole C; Waters, Clarice N; Estabrooks, Paul A; You, Wen; Zoellner, Jamie M

    2015-03-26

    Interest has increased in understanding the types and healthfulness of restaurant foods for children, particularly in disadvantaged areas. The purpose of this community-based participatory research study was to describe the quality of restaurant food offered to children in a health-disparate region in Virginia and North Carolina and to determine if the availability of healthy foods differed by location (rural, urban) or by the predominant race (black, white, mixed race) of an area's population. Restaurants offering a children's menu in the 3 counties in Virginia and North Carolina that make up the Dan River Region were identified by using state health department records. Research assistants reviewed menus using the Children's Menu Assessment (CMA), a tool consisting of 29 scored items (possible score range, -4 to 21). Scores were calculated for each restaurant. We obtained information on the predominant race of the population at the block group level for all counties from 2010 US Census data. For the 137 restaurants studied, mean CMA scores were low (mean, 1.6; standard deviation [SD], 2.7), ranging from -4 to 9 of 21 possible points. Scores were lowest for restaurants in the predominantly black block groups (mean, 0.2; SD, 0.4) and significantly different from the scores for restaurants in the predominantly white (mean, 1.4; SD, 1.6) and mixed-race block groups (mean, 2.6; SD, 2.4) (F = 4.3; P < .05). Children's menus available in the Dan River Region lack healthy food options, particularly in predominantly black block groups. These study findings can contribute to regional efforts in policy development or environmental interventions for children's food quality by the community-based participatory research partnership and help local stakeholders to determine possible strategies and solutions for improving local food options for children.

  20. Lack of Healthy Food Options on Children’s Menus of Restaurants in the Health-Disparate Dan River Region of Virginia and North Carolina, 2013

    Science.gov (United States)

    Olive, Nicole C.; Waters, Clarice N.; Estabrooks, Paul A.; You, Wen; Zoellner, Jamie M.

    2015-01-01

    Introduction Interest has increased in understanding the types and healthfulness of restaurant foods for children, particularly in disadvantaged areas. The purpose of this community-based participatory research study was to describe the quality of restaurant food offered to children in a health-disparate region in Virginia and North Carolina and to determine if the availability of healthy foods differed by location (rural, urban) or by the predominant race (black, white, mixed race) of an area’s population. Methods Restaurants offering a children’s menu in the 3 counties in Virginia and North Carolina that make up the Dan River Region were identified by using state health department records. Research assistants reviewed menus using the Children’s Menu Assessment (CMA), a tool consisting of 29 scored items (possible score range, −4 to 21). Scores were calculated for each restaurant. We obtained information on the predominant race of the population at the block group level for all counties from 2010 US Census data. Results For the 137 restaurants studied, mean CMA scores were low (mean, 1.6; standard deviation [SD], 2.7), ranging from −4 to 9 of 21 possible points. Scores were lowest for restaurants in the predominantly black block groups (mean, 0.2; SD, 0.4) and significantly different from the scores for restaurants in the predominantly white (mean, 1.4; SD, 1.6) and mixed-race block groups (mean, 2.6; SD, 2.4) (F = 4.3; P < .05). Conclusion Children’s menus available in the Dan River Region lack healthy food options, particularly in predominantly black block groups. These study findings can contribute to regional efforts in policy development or environmental interventions for children’s food quality by the community-based participatory research partnership and help local stakeholders to determine possible strategies and solutions for improving local food options for children. PMID:25811495

  1. Neutron flux determination at the horizontal beam tubes of the Rossendorf research reactor RFR

    International Nuclear Information System (INIS)

    Stephan, I.; Schneider, B.; Boehmer, B.; Boehme, K.; Mehner, H.C.

    1992-01-01

    To characterize the irradiation possibilities of the RFR, in particular the values concerning the steel tubes, diffusion- and transport-theoretical calculations have been made and supported by experiments. The measurements also served to provide secured data for the experts on the ad hoc committee ''Rossendorf Research Reactor'' to assess the neutron situation of German research reactors. (orig.) [de

  2. The NCSU [North Carolina State Univ.] freon PWR [pressurized water reactor] loop

    International Nuclear Information System (INIS)

    Caves, J.R.; Doster, J.M.; Miller, G.D.; Wehring, B.W.; Turinsky, P.J.

    1989-01-01

    The nuclear engineering department at North Carolina State University has designed and constructed an operating scale model of a pressurized water reactor (PWR) nuclear steam supply system (NSSS). This facility will be used for education, training, and research. The loop uses electric heaters to simulate the reactor core and Freon as the primary and secondary coolant. Viewing ports at various locations in the loop allow the students to visualize flow regimes in normal and off-normal operating conditions. The objective of the design effort was to scale the thermal-hydraulic characteristics of a two-loop Westinghouse NSSS. Provisions have been made for the simulation of various abnormal occurrences. The model is instrumented in much the same manner as the actual NSSS. Current research projects using the loop include the development of adaptive expert systems to monitor the performance of the facility, diagnose mechanical faults, and to make recommendations to operators for mitigation of accidents. This involves having thermal-hydraulics and core-physics simulators running faster than real time on a mini-supercomputer, with operating parameters updated by communication with the data acquisition and control computer. Further opportunities for research will be investigated as they arise

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

    International Nuclear Information System (INIS)

    Shure, D.J.; Gottschalk, M.R.

    1975-01-01

    Cesium-137 dynamics were studied in a blackwater creek which had received production reactor effluents from the Savannah River Plant in South Carolina. Most 137 Cs 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 137 Cs 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 137 Cs 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 137 Cs 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 137 Cs concentrations despite a 10-year absence of reactor operations. Radiocesium levels are apparently being maintained through long-term 137 Cs cycling in the upstream reservoir and surrounding flood plain forest systems. Rainfall and other physical processes influence the seasonal 137 Cs fluctuations in stream components. (auth)

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

    International Nuclear Information System (INIS)

    Rajendran, S.; Ramakumar, M.S.

    1996-01-01

    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)

  5. The cracking of pressure tubes in the Pickering reactor

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.

    1978-01-01

    Small cracks in 17 of the 390 pressure tubes in Unit 3 of the 2056 MW (electrical) Pickering Generating Station and of 52 tubes in Unit 4, resulted in each of these units being out of service for many months. The cracks originated at areas of extremely high residual tensile stress produced by improper positioning of the rolling tool used during construction to join the pressure tube to its end-fitting. The mechanism of failure was delayed hydrogen cracking. (author)

  6. A Regional View of the Margin: Salmonid Abundance and Distribution in the Southern Appalachian Mountains of North Carolina and Virginia

    Science.gov (United States)

    Patricia A. Flebbe

    1994-01-01

    In the southern Appalachian Mountains, native brook trout Salvelinus fontinalis and introduced rainbow trout Oncorhynchus mykiss and brown trout Salmo trutta are at the southern extremes of their distributions, an often overlooked kind of marginal habitat. At a regional scale composed of the states of Virginia...

  7. 76 FR 18216 - Dominion Virginia Power/North Carolina Power; Notice of Availability of Shoreline Management Plan...

    Science.gov (United States)

    2011-04-01

    .... Application Type: Revised Shoreline Management Plan. b. Project No.: 2778-062. c. Date Filed: December 29... of the Roanoke Rapids and Gaston Hydroelectric Project, has filed a revised Shoreline Management Plan... Power/North Carolina Power; Notice of Availability of Shoreline Management Plan Update for the Shoshone...

  8. Analysis of defect tubes of fast reactor heat exchanger

    International Nuclear Information System (INIS)

    Rukhlyada, N.Ya.

    2014-01-01

    The experimental Auger electron spectroscopy and X-ray diffraction microanalysis data of laboratory investigations of defect tubes of heat exchanger with sodium coolant are presented. Element distribution through depth of corrosion layers form on the side of coolant (sodium) and on the surface contacting with steam in heat exchanger tube is studied. Sodium presence through all thickness of the tube is determined. It is shown that treatment of 12Cr18N9 steel surface by plasma pulses decreases intergranular corrosion susceptibility. It is related with structural changes of surface layer (∼ 20 μm), its enrichment by chromium and formation of chromium oxide protective film [ru

  9. Novel electromagnetic technique for repositioning of coolant tube spacers in CANDU nuclear reactors

    Science.gov (United States)

    Dableh, Joseph H.

    1986-06-01

    A novel electromagnetic technique to reposition the coolant tube spacers in the fuel channels of CANDU nuclear reactors was successfully developed in the fall of 1983 at Ontario Hydro Research Division. The need to reposition dislocated spacers in noncommissioned reactors was discovered subsequent to the rupture of a pressure tube in one reactor at the Pickering Nuclear Generator Station in Ontario. A contributing factor to the failure of the tube was the fact that the annular spacers (garter springs), used to maintain the coaxial configuration between the pressure tube and its surrounding calandria tube, had been displaced longitudinally for a number of years. Subsequent to this finding, it was discovered that a number of garter springs in noncommissioned nuclear reactors were displaced due to vibration induced by various sources during the construction stage. Since the garter springs are not directly accessible by mechanical means, extensive dismantling of the fuel channels would have been necessary to reposition the springs in their designated locations. This paper describes a novel method to reposition the garter springs without dismantling the fuel channels. The method consists of exerting a force on the springs in the direction of the required displacement by applying a large electromagnetic impulse (generated by a 200-kJ capacitor bank) to a drive coil inserted into the pressure tube opposite the spacer. The repositioning of displaced garter springs in five new reactors in Ontario has been carried out successfully in 1984. The saving in reactor repair cost, interest charges, and replacement energy cost was on the order of hundreds of millions of dollars. Equally large benefits and savings will be realized if the need to use this technique in commissioned reactors arises. Also, the related development of strong compact coils and low-resistance pulse power cable have significant implications and advantages in various other applications related to the pulse

  10. Steam generator tube failures: experience with water-cooled nuclear power reactors during 1976

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1978-02-01

    A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author)

  11. 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.

  12. Studies on the instrumentation of a beam-tube medium flux reactor

    International Nuclear Information System (INIS)

    Axmann, A.; Pollet, J.L.; Queudot, J.

    1979-01-01

    In the years 1977/78, the ad hoc commitee for medium-flux reactor development of the Federal Ministry for Research and Technology developed constructional concepts for a medium-flux reactor to be utilized by beam tube experiments. The HMI has elaborated contributions for discussions of the subject of instrumentation, in particular for experiments in solid state physics. These contributions are contained in the report. (orig./RW) [de

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

    International Nuclear Information System (INIS)

    Sudjatmi KA

    2013-01-01

    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)

  14. Utilization of Stop-flow Micro-tubing Reactors for the Development of Organic Transformations.

    Science.gov (United States)

    Toh, Ren Wei; Li, Jie Sheng; Wu, Jie

    2018-01-04

    A new reaction screening technology for organic synthesis was recently demonstrated by combining elements from both continuous micro-flow and conventional batch reactors, coined stop-flow micro-tubing (SFMT) reactors. In SFMT, chemical reactions that require high pressure can be screened in parallel through a safer and convenient way. Cross-contamination, which is a common problem in reaction screening for continuous flow reactors, is avoided in SFMT. Moreover, the commercially available light-permeable micro-tubing can be incorporated into SFMT, serving as an excellent choice for light-mediated reactions due to a more effective uniform light exposure, compared to batch reactors. Overall, the SFMT reactor system is similar to continuous flow reactors and more superior than batch reactors for reactions that incorporate gas reagents and/or require light-illumination, which enables a simple but highly efficient reaction screening system. Furthermore, any successfully developed reaction in the SFMT reactor system can be conveniently translated to continuous-flow synthesis for large scale production.

  15. Steam generator tube rupture in an experimental facility scaled from a pressurized water reactor

    International Nuclear Information System (INIS)

    Loomis, G.G.

    1984-09-01

    Results from an experimental investigation of steam generator tube rupture in the Semiscale Mod-2B system are presented. From the experimental results, the characteristic system response signature for a wide range of number of tubes ruptured has been described. The tube rupture was assumed to occur during normal full power operation (15.6 MPa system pressure, 37 0 K core differential temperature). In addition, recovery scenarios involving operator actions were examined. The recovery scenarios included use of pressurizer auxiliary spray and internal heaters, steam generator feed and steam, primary feed and bleed, and main cooling pump operation. Recovery scenarios suggested by typical US pressurized water reactor emergency operating procedures were followed

  16. Integrity evaluation for steam generator tube of system integrated modular advanced reactor

    International Nuclear Information System (INIS)

    Kim, J. S.; Jin, T. E.; Jeong, M. J.; Choi, Y. H.; Jeo, J. C.

    2003-01-01

    In this study, the structural integrity for SG tube of system integrated modular advanced reactor, which is subjected to dominant external pressure as well as helical type, is evaluated using the commercial finite element package ABAQUS and the American petrochemical industry code API 579 Appendix B. First of all, the crack behavior under the assumption of local heating is assessed using ABAQUS. And, the buckling behavior of tube with 40% wall thinning is assessed using API 579 Appendix B. As a result, it is found that the crack closure phenomenon occurs under external pressure and the buckling doesn't occur even if 40% wall thinning exists in tube

  17. Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2012-02-15

    Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

  18. Conceptual design of a bayonet-tube steam generator for the ALFRED lead-cooled reactor

    International Nuclear Information System (INIS)

    Damiani, Lorenzo; Montecucco, Massimo; Pini Prato, Alessandro

    2013-01-01

    Highlights: • Conceptual design of a bayonet-tube steam generator for a lead-cooled reactor demonstrator. • Steady-state simulations effected through RELAP 5. • Performance evaluation of the steam generator for different configurations of the bayonet tubes. - Abstract: The present paper is centred on the design of a bayonet tube steam generator, fundamental part of an innovative lead-cooled fast nuclear reactor (LFR). The construction of the LFR is the main objective of the European project named LEADER, of which Ansaldo Nucleare is an important member. The steam generator described in this paper is expected to be installed in a 300 MW thermal power demonstrator plant, named ALFRED. The investigations carried out in this work through the RELAP 5 code have first faced the sizing of the single bayonet tube and then the design of the whole heat exchanger. The configurations of the four coaxial tubes composing the single bayonet, the length of the bayonets and the materials employed have been investigated; the final heat exchanger configuration provides 510 bayonet tubes of 6 m active length with a thermal insulation between the inner descending tube and the rising annulus, assured by a special extremely insulating paint. The whole steam generator has shown its capability to reach the required exchanged power of 37.5 MW th , providing as output dry superheated steam at the desired temperature of 450 °C

  19. 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.

  20. The Caltech Photooxidation Flow Tube reactor: design, fluid dynamics and characterization

    Science.gov (United States)

    Huang, Yuanlong; Coggon, Matthew M.; Zhao, Ran; Lignell, Hanna; Bauer, Michael U.; Flagan, Richard C.; Seinfeld, John H.

    2017-03-01

    Flow tube reactors are widely employed to study gas-phase atmospheric chemistry and secondary organic aerosol (SOA) formation. The development of a new laminar-flow tube reactor, the Caltech Photooxidation Flow Tube (CPOT), intended for the study of gas-phase atmospheric chemistry and SOA formation, is reported here. The present work addresses the reactor design based on fluid dynamical characterization and the fundamental behavior of vapor molecules and particles in the reactor. The design of the inlet to the reactor, based on computational fluid dynamics (CFD) simulations, comprises a static mixer and a conical diffuser to facilitate development of a characteristic laminar flow profile. To assess the extent to which the actual performance adheres to the theoretical CFD model, residence time distribution (RTD) experiments are reported with vapor molecules (O3) and submicrometer ammonium sulfate particles. As confirmed by the CFD prediction, the presence of a slight deviation from strictly isothermal conditions leads to secondary flows in the reactor that produce deviations from the ideal parabolic laminar flow. The characterization experiments, in conjunction with theory, provide a basis for interpretation of atmospheric chemistry and SOA studies to follow. A 1-D photochemical model within an axially dispersed plug flow reactor (AD-PFR) framework is formulated to evaluate the oxidation level in the reactor. The simulation indicates that the OH concentration is uniform along the reactor, and an OH exposure (OHexp) ranging from ˜ 109 to ˜ 1012 molecules cm-3 s can be achieved from photolysis of H2O2. A method to calculate OHexp with a consideration for the axial dispersion in the present photochemical system is developed.

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

    International Nuclear Information System (INIS)

    Prepechal, J.; Sladky, J.

    1974-01-01

    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.)

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    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.

  4. 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)

    1997-04-01

    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

  5. The management of severe accidents in modern pressure tube reactors

    International Nuclear Information System (INIS)

    Popov, N.K.; Santamaura, P.; Blahnik, C.; Snell, V.G.; Duffey, R.B.

    2007-01-01

    Advanced new reactor designs resist severe accidents through a balance between prevention and mitigation. This balance is achieved by designing to ensure that such accidents are very rare; and by limiting core damage progression and releases from the plant in the event of such rare accidents. These design objectives are supported by a suitable combination of probabilistic safety analysis, engineering judgment and experimental and analytical study. This paper describes the approach used for the Advanced CANDU Reactor TM -1000 (ACR-1000) design, which includes provisions to both prevent and mitigate severe accidents. The paper describes the use of PSA as a 'design assist' tool; the analysis of core damage progression pathways; the definition of the core damage states; the capability of the mitigating systems to stop and control severe accident events; and the severe accident management opportunities for consequence reduction. (author)

  6. Simulation of Reforming Reactor Tube: Quantifying Catalyst Pellet's Effectiveness Factor

    OpenAIRE

    Da Cruz, Flavio Eduardo

    2016-01-01

    In this work, a consistent mathematical model to simulate a spherical catalytic pellet and a Packed-Bed Reactor (PBR) is develop. The Dusty Gas Model (DGM) is applied to the calculation of the diffusive fluxes in the porous media. Simulations are executed considering hydrogen production from steam methane reforming. Species’ diffusivities are calculated using data from literature as well as the values for tortuosity and porosity. The pellet simulation is performed considering mass, species, m...

  7. Carbon nanotubes: from nano test tube to nano-reactor.

    Science.gov (United States)

    Khlobystov, Andrei N

    2011-12-27

    Confinement of molecules and atoms inside carbon nanotubes provides a powerful strategy for studying structures and chemical properties of individual molecules at the nanoscale. In this issue of ACS Nano, Allen et al. explore the nanotube as a template leading to the formation of unusual supramolecular and covalent structures. The potential of carbon nanotubes as reactors for synthesis on the nano- and macroscales is discussed in light of recent studies.

  8. 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.

  9. Simulation of Water Quality in the Tull Creek and West Neck Creek Watersheds, Currituck Sound Basin, North Carolina and Virginia

    Science.gov (United States)

    Garcia, Ana Maria

    2009-01-01

    A study of the Currituck Sound was initiated in 2005 to evaluate the water chemistry of the Sound and assess the effectiveness of management strategies. As part of this study, the Soil and Water Assessment Tool (SWAT) model was used to simulate current sediment and nutrient loadings for two distinct watersheds in the Currituck Sound basin and to determine the consequences of different water-quality management scenarios. The watersheds studied were (1) Tull Creek watershed, which has extensive row-crop cultivation and artificial drainage, and (2) West Neck Creek watershed, which drains urban areas in and around Virginia Beach, Virginia. The model simulated monthly streamflows with Nash-Sutcliffe model efficiency coefficients of 0.83 and 0.76 for Tull Creek and West Neck Creek, respectively. The daily sediment concentration coefficient of determination was 0.19 for Tull Creek and 0.36 for West Neck Creek. The coefficient of determination for total nitrogen was 0.26 for both watersheds and for dissolved phosphorus was 0.4 for Tull Creek and 0.03 for West Neck Creek. The model was used to estimate current (2006-2007) sediment and nutrient yields for the two watersheds. Total suspended-solids yield was 56 percent lower in the urban watershed than in the agricultural watershed. Total nitrogen export was 45 percent lower, and total phosphorus was 43 percent lower in the urban watershed than in the agricultural watershed. A management scenario with filter strips bordering the main channels was simulated for Tull Creek. The Soil and Water Assessment Tool model estimated a total suspended-solids yield reduction of 54 percent and total nitrogen and total phosphorus reductions of 21 percent and 29 percent, respectively, for the Tull Creek watershed.

  10. Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor (SFR)

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock

    2013-01-01

    To achieve manufacturing technology of the fuel cladding tube in order to keep pace with the predetermined schedule in developing SFR fuel, KAERI has launched in developing fuel cladding tube in cooperation with a domestic steelmaking company. After fabricating medium-sized 1.1 ton HT9 ingot, followed by the multiple processes of hot and cold working, preliminary samples of HT9 seamless cladding tube having 7.4mm in outer diameter, 0.56mm in thickness, and 3m in length were fabricated. The objective of this study is to summarize the brief development status of the HT9 cladding tubes. Mechanical properties like axial tension, biaxial burst, pressurized creep and sodium compatibility of the cladding tubes were carried out to set up the performance evaluation technology to test the prototype FMS cladding tube which is going to be manufactured in next stage. As a part of developing fuel cladding for the Sodium-cooled Fast Reactor (SFR), preliminary HT9 cladding tube was fabricated in cooperation with a domestic steelmaking company. Microstructure as well as mechanical tests like axial tensile test, biaxial burst test, and pressurized creep test of the fuel cladding were carried out. Performance of the domestic HT9 tube was revealed to be similar in the previously fabricated foreign HT9 tube. Further prototype FMS cladding tube is going to be manufactured in next year based on this experience. Various test items like mechanical test, sodium compatibility test, microstructural analysis, basic property, cladding performance under transient situation, and performance under ion and neutron irradiation are going be performed in the future to set up the relevant technology for the licensing of the SFR cladding tube

  11. The development of octagon Zr-4 alloy tube for heating reactors

    International Nuclear Information System (INIS)

    Yang Fanglin; Yang Yingli; Wang Guangshen

    1989-10-01

    The asymmetrical octagon Zr-4 alloy tubes which are used for fuel assembly in the heating reactor have been developed. The thickness of tube wall is 1.5 mm and the length is 1725 mm. The long side of the octagon is 138.7 0.3 +0.2 mm, the short side is 93.1 ± 0.1 mm. To manufacture these tubes a stretch draw forming processing method is adopted. The process is divided into two phases. In the first phase, a short draw mould is used to stretch the Zr-4 alloy tube. In the second phase, a long draw mould, its length is equal to the end-produt length, is used to complete the final processing. The size accuracy and repeatability of this method are excellent and can fully meet the design requirements

  12. Development of out-of-core concepts for a supercritical-water, pressure-tube reactor

    International Nuclear Information System (INIS)

    Diamond, W.T.

    2010-01-01

    One of the Generation IV programs at Chalk River Laboratories has as its prime focus the development of out-of-core concepts for the SuperCritical Water (SCW) pressure tube reactor under development in Canada. A number of technical issues associated with the interface of out-of-core components and the pressure tubes of a SCW pressure tube reactor are being investigated. This article focuses on several aspects of out-of-core components and layouts, building upon concepts that have been developed during the past few years. The efforts are strongly focused on concepts for a fuel channel that can be fabricated with the tight lattice pitch (typically 230 to 250 mm) that may be required for some applications such as utilization of a thorium fuel cycle. It is not practical to adapt concepts with a tight lattice pitch while using the thicker materials required for the higher temperatures and pressures required for supercritical operation. A change in lattice pitch or configuration is required to accommodate the component size increases. This presentation will cover a number of new concepts developed to produce feeders and end fittings for the harsh conditions of a SCW pressure tube reactor. These components are then developed into conceptual models of a Gen IV pressure tube reactor mounted in both horizontal and vertical orientations. Full 3-D solid models of both concepts will be demonstrated as well as a 1/10th-scale model of one face of a horizontal concept that has been built from components made with a 3-D printer. (author)

  13. Thermal hydraulic stability in a pressure tube nuclear reactor

    International Nuclear Information System (INIS)

    Villani, A.; Ravetta, R.; Mansani, L.

    1986-01-01

    The CIRENE plant which will undergo preoperational tests in the near future is equipped with a 40 MW(e) Heavy Water moderated Boiling Light Water cooled Reactor (HWBLWR); at the start-up and up to about 30 % of nominal power, the necessary low coolant density is obtained injecting into the core a mixture of liquid and steam. To verify the thermal-hydraulic stability of the plant in this situation, tests have been carried out in a facility simulating two full scale power channels; the system stability has been confirmed in the reference conditions, and is not reduced by even a significant reduction of the liquid flowrate, where a decrease in liquid temperature has some negative effect and steam flowrate has a small influence. (author)

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

    Science.gov (United States)

    McDermott, D.J.; Schrader, K.J.; Schulz, T.L.

    1994-05-03

    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.

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

    Science.gov (United States)

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    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. Effect of microstructural evolution on in-reactor creep of Zr-2.5Nb tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, YoungSuk [Korea Atomic Energy Research Institute, Zirconium Group, P.O. Box 105, Yusong, Daejon 305-353 (Korea, Republic of)]. E-mail: yskim1@kaeri.re.kr; Im, KyungSoo [Korea Atomic Energy Research Institute, Zirconium Group, P.O. Box 105, Yusong, Daejon 305-353 (Korea, Republic of); Cheong, YongMoo [Korea Atomic Energy Research Institute, Zirconium Group, P.O. Box 105, Yusong, Daejon 305-353 (Korea, Republic of); Ahn, SangBok [Korea Atomic Energy Research Institute, Zirconium Group, P.O. Box 105, Yusong, Daejon 305-353 (Korea, Republic of)

    2005-11-15

    Dislocation density, decomposition of the {beta}-Zr phase and diametral creep were examined as a function of the location of the Zr-2.5Nb tube irradiated in the Wolsong Unit 1 for 9.3 effective full power years (EFPYs). The maximum a-dislocation density occurred at the inlet part of the irradiated Zr-2.5Nb tube exposed to the lowest temperature while the outlet part of the tube exposed to the higher temperature had the higher extent of decomposition of the {beta}-Zr phase and the maximum diametral creep. Thus, it is concluded that in-reactor creep of the Zr-2.5Nb tube is not related to the dislocation density but governed by the Nb concentration of the {alpha}-Zr grains caused by thermal decomposition of the {beta}-Zr phase. Supplementary creep tests on the Zr-2.5Nb sheets with different Nb contents in the {beta}-Zr phase provide supportive evidence to this conclusion. The acceleration of the in-reactor creep of the Zr-2.5Nb tubes is suggested after a long-term operation.

  18. Water-quality assessment of the Albemarle-Pamlico drainage basin, North Carolina and Virginia; characterization of suspended sediment, nutrients, and pesticides

    Science.gov (United States)

    Harned, Douglas; McMahon, Gerard; Spruill, T.B.; Woodside, M.D.

    1995-01-01

    The 28,000-square-mile Albemarle-Pamlico drainage basin includes the Roanoke, Dan, Chowan Tar, and Neuse Rivers. The basin extends through four physiographic provinces in North Carolina and Virginia-Valley and Ridge, Blue Ridge, Piedmont and Coastal Plain. The spatial and temporal trends in ground-water and riverine water quality in the study area were characterized by using readily available data sources The primary data sources that were used included the U.S. Geological Survey's National Water Data Storage and Retrieval System (WATSTORE) database, the U.S. Environmental Protection Agency's Storage and Retrieval System (STORET) database, and results of a few investigations of pesticide occurrence. The principal water-quality constituents examined were suspended sediment, nutrients, and pesticides. The data examined generally spanned the period from 1950 to 1993. The only significant trends in suspended sediment were detected at three Chowan River tributary sites which showed long-term decreases. Suspended- and total-solids concentrations have decreased throughout the Albemarle-Pamlico drainage basin. The decreases are probably a result of (1) construction of new lakes and ponds in the basin, which trap solids, (2) improved agricultural soil management, and (3) improved wastewater treatment. Nutrient point sources are much less than nonpoint nutrient sources at the eight NASQAN basins examined for nutrient loads. The greatest nitrogen inputs are associated with crop fertilizer and biological nitrogen fixation by soybeans and peanuts, whereas atmospheric and animal-related nitrogen inputs are comparable in magnitude. The largest phosphorus inputs are associated with animal wastes. The most commonly detected pesticides in surface water in the STORET database were atrazine and aldrin.Intensive organonitrogen herbicide sampling of Chicod Creek in 1992 showed seasonal variations in pesticide concentration. The most commonly detected herbicides were atrazine, alachlor

  19. Analysis of effects of calandria tube uncovery under severe accident conditions in CANDU reactors

    International Nuclear Information System (INIS)

    Rogers, J.T.; Currie, T.C.; Atkinson, J.C.; Dick, R.

    1983-01-01

    A study is being undertaken for the Atomic Energy Control Board to assess the thermal and hydraulic behaviour of CANDU reactor cores under accident conditions more severe than those normally considered in the licensing process. In this paper, we consider the effects on a coolant channel of the uncovery of a calandria tube by moderator boil-off following a LOCA in a Bruce reactor unit in which emergency cooling is ineffective and the moderator heat sink is impaired by the failure of the moderator cooling system. Calandria tube uncovery and its immediate consequences, as described here, constitute only one part of the entire accident sequence. Other aspects of this sequence as well as results of the analysis of the other accident sequences studied will be described in the final report on the project and in later papers

  20. Remote-controlled television for locating leaking tubes in pressurized-water reactor steam generators

    International Nuclear Information System (INIS)

    Cormault, P.; Denis, J.

    1978-01-01

    The Scarabee system is designed for observation of the tubes in water boxes of pressurized-water reactor nuclear-power-station steam generators. It consists essentially of a camera and a projector used as a marker, both of which swivel freely. The whole unit is housed in a water-tight container which can easily be decontaminated. Remote control of camera and marker movement is carried out from a console. (author)

  1. Experience in quality assurance of alloy D9 clad tubes for Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Kapoor, K.; Prahlad, B.

    2012-01-01

    Stainless Steel Alloy D9 is the material for cladding in various sub-assemblies of Prototype Fast Breeder Reactor (PFBR). The fabrication, inspection, testing and supply of the clad tubes for the first core of PFBR is nearly completed. The paper also compares the specification requirements and the achieved results for some of the critical aspects which is arrived after completing supply against the first core requirement

  2. Design and use of the ORNL HFIR [High Flux Isotope Reactor] pneumatic tube irradiation systems

    International Nuclear Information System (INIS)

    Dyer, F.F.; Emery, J.F.; Robinson, L.; Teasley, N.A.

    1987-01-01

    A second pneumatic tube that was recently installed in the High Flux Isotope Reactor for neutron activation analysis is described. Although not yet tested, the system is expected to have a thermal neutron flux of about 1.5 x 10 14 cm -2 s -1 . A delayed neutron counter is an integral part of the pneumatic tube, and all of the hardware is present to enable automated use of the counter. The system is operated with a Gould programmable controller that is programmed with an IBM personal computer. Automation of any mode of operation, including the delayed neutron counter, will only require a nominal amount of software development. Except for the lack of a hot cell, the irradiation facility has all of the advantageous features of an older pneumatic tube that has been in operation for 17 years. The design of the system and some applications and methods of operation are described

  3. Reactor

    International Nuclear Information System (INIS)

    Evans, R.M.

    1976-01-01

    Disclosed is a neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch. 1 claim, 16 figures

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

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1983-08-01

    Heating tests using 53 wt % U 3 O 8 -Al pellets show that an exothermic reaction occurs between 875 and 1000 0 C and takes 10 to 20 seconds to reach maximum temperature. The maximum temperature is a function of particle size of the U 3 O 8 with large particles exhibiting lower peak temperatures. The calculated energy release was 123 cal/g of U 3 O 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 3 O 8 -Al reaction is not an important energy source. The compressive and tensile strengths of U 3 O 8 tubes above 660 0 C are low. In compression, sections with 2 psi average axial stress failed at 917 0 C, while sections with 7 psi failed at 669 0 C. Tubes with U-Al alloy cores failed at about 670 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

  5. Non-destructive evaluation of stream generator tubes and pressure tubes from the PHWR reactors, using the rotating magnetic field method

    International Nuclear Information System (INIS)

    Premel, D.; Placko, D.; Grimberg, R.; Savin, A.

    2001-01-01

    This work presents a new type of eddy current transducer with a rotating magnetic field devoted to the inspection of steam generator tubes and pressure tubes from the PHWR reactors. A theoretical model has been developed that permits the calculations of the emf induced in the reception coils in the presence of the copper or magnetite deposits, anti-vibration railing and garter springs. (authors)

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

    International Nuclear Information System (INIS)

    Roegler, Hans Joachim; Feltes, Wolfgang

    1998-01-01

    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)

  7. 78 FR 16816 - Television Broadcasting Services; Hampton-Norfolk, Virginia; Norfolk, Virginia-Elizabeth City...

    Science.gov (United States)

    2013-03-19

    ... FEDERAL COMMUNICATIONS COMMISSION 47 CFR Part 73 [MB Docket No. 11-139; RM-11636; DA 13-258] Television Broadcasting Services; Hampton-Norfolk, Virginia; Norfolk, Virginia-Elizabeth City, North Carolina... modify its television station, WHRO-TV's license to specify Norfolk, Virginia-Elizabeth City, North...

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

    International Nuclear Information System (INIS)

    Cvitanovic, M.; Zado, V.

    1996-01-01

    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)

  9. Power level effects on thorium-based fuels in pressure-tube heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bromley, B.P.; Edwards, G.W.R., E-mail: blair.bromley@cnl.ca [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Sambavalingam, P. [Univ. of Ontario Inst. of Technology, Oshawa, Ontario (Canada)

    2016-06-15

    Lattice and core physics modeling and calculations have been performed to quantify the impact of power/flux levels on the reactivity and achievable burnup for 35-element fuel bundles made with Pu/Th or U-233/Th. The fissile content in these bundles has been adjusted to produce on the order of 20 MWd/kg burnup in homogeneous cores in a 700 MWe-class pressure-tube heavy water reactor, operating on a once-through thorium cycle. Results demonstrate that the impact of the power/flux level is modest for Pu/Th fuels but significant for U-233/Th fuels. In particular, high power/flux reduces the breeding and burnup potential of U-233/Th fuels. Thus, there may be an incentive to operate reactors with U-233/Th fuels at a lower power density or to develop alternative refueling schemes that will lower the time-average specific power, thereby increasing burnup.(author)

  10. The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

    International Nuclear Information System (INIS)

    Zheng Youqi; Zu Tiejun; Wu Hongchun; Cao Liangzhi; Yang Chao

    2012-01-01

    In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.

  11. Pressure tube reactors and a sustainable energy future: the ultra development path

    International Nuclear Information System (INIS)

    Duffey, R.

    2008-01-01

    Nuclear energy must be made available, freely and readily, to help meet world energy needs, concerns over energy price and security of supply, and alleviating the uncertainties over potential climate change. The perspective offered here is a model for others to consider, adopting and adapting using whatever elements fit their own strategies and needs. The underlying philosophy is to retain flexibility in the reactor development, deployment and fuel cycle, while ensuring the principle that customer, energy market, safety, non-proliferation and sustainability needs are all addressed. Canada is the world's largest exporter of uranium, providing about one-third of the world supply for nuclear power reactors. Pressure tube reactors (PTRs), of which CANDU, is a prime example, have a major role to play in a sustainable energy future. The inherent fuel cycle flexibility of the PTR offers many technical, resource and sustainability and economic advantages over other reactor technologies and is the subject of this paper. The design evolution and development intent is to be consistent with improved or enhanced safety, licensing and operating limits, global proliferation concerns, and waste stream reduction, thus enabling sustainable energy futures. The limits are simply those placed by safety, economics and resource availability. (author)

  12. Pressure tube reactors and a sustainable energy future: the ultra development path

    International Nuclear Information System (INIS)

    Duffey, R.

    2008-01-01

    Nuclear energy must be made available, freely and readily, to help meet world energy needs, concerns over energy price and security of supply, and alleviating the uncertainties over potential climate change. The perspective offered here is a model for others to consider, adopting and adapting using whatever elements fit their own strategies and needs. The underlying philosophy is to retain flexibility in the reactor development, deployment and fuel cycle, while ensuring the principle that customer, energy market, safety, non-proliferation and sustainability needs are all addressed. Canada is the world's largest exporter of uranium, providing about one-third of the world supply for nuclear power reactors. Pressure tube reactors (PTRs), of which CANDU is a prime example, have a major role to play in a sustainable energy future. The inherent fuel cycle flexibility of the PTR offers many technical, resource and sustainability, and economic advantages over other reactor technologies and is the subject of this paper. The design evolution and development intent is to be consistent with improved or enhanced safety, licensing and operating limits, global proliferation concerns, and waste stream reduction, thus enabling sustainable energy futures. The limits are simply those placed by safety, economics and resource availability. (author)

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

    International Nuclear Information System (INIS)

    Urbanic, V.F.; Warr, B.D.; Manolescu, A.; Chow, C.K.; Shanahan, M.W.

    1989-01-01

    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/dm 2 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)

  14. Applications of Ground Penetrating Radar for Hydrogeologic Characterization at the P Reactor Area, Savannah River Site, South Carolina

    Science.gov (United States)

    Cameron, A. E.; Knapp, C. C.; Addison, A.; Waddell, M.

    2006-12-01

    Ground Penetrating Radar (GPR) techniques were implemented at Savannah River Site (SRS), South Carolina, in order to develop new approaches for hydrogeophysical characterization in heterogeneous environments. The study site is the P Reactor Area located within the Upper Atlantic Coastal Plain, with clastic sediments ranging from Late Cretaceous to Miocene age. Lithologies consist of sand, clayey sand, and clay with minor amounts of calcareous minerals. Increasing interest in the P Reactor Area in recent years is the result of the presence of several contaminant plumes including trichloroethylene (TCE) that originates from the northwest section of the reactor facility and discharges into nearby Steel Creek. Here, we present the results from subsurface characterization using the GPR technique involving the PulseEKKO 100 GPR system with 50, 100, and 200 MHz antennas. Data acquisition included (1) several regional 2D common-offset GPR transects for general stratigraphic characterization, (2) a series of common-mid point (CMP) profiles for velocity estimation, (3) a set of vertical radar profiles (VRP) at an existing borehole in the vicinity of the study area, and (4) a 3D GPR survey for detailed subsurface lithostratigraphic characterization collected with the 50 MHz antenna. The ongoing GPR processing will map key dielectric interfaces from the ground surface to ~20 meters, and will be integrated with co-located surface and vertical seismic reflection data as well as with lithologic information obtained from Cone Penetrometer Technology (CPT) tests performed at the study area. This lithosptratigraphic zonation will provide the framework for subsequent hydrological parameter estimation, which will be performed using borehole hydrogeological and crosshole seismic and GPR methods. This research was supported by the Office of Science (BER), U.S. Department of Energy, Grant No. DE-FG02-06ER64210.

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

    International Nuclear Information System (INIS)

    Peacock, H.B.

    1984-01-01

    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 3 O 8 -aluminum cermets. Above the aluminum melting point, U 3 O 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 therefore 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 3 O 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 0 C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U 3 O 8 -aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 660 0 C. In tension, sections failed at about the aluminum melting point. In compression with 2 psi average axial stress, failure occurred at 917 0 C, while 7 psi average axial stress produced failure at 669 0 C. (author)

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

    International Nuclear Information System (INIS)

    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 3 O 8 -aluminum cermets. Above the aluminum melting point, U 3 O 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 3 O 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 0 C in air without a violent effect. The heat of reaction was approximately 123 cal/g of U 3 O 8 -aluminum fuel. Tensile and compressive strength of the fuel tube section is low above 660 0 C. In tension, sections failed at about the aluminum melting point. In compression with 2-psi average axial stress, failure occurred at 917 0 C, while 7 psi average axial stress produced failure at 669 0 C

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

    International Nuclear Information System (INIS)

    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 x 10 8 n/cm 2 · s. The fast neutron and gamma radiation KERMA factors are 10 x 10 -11 cGy·cm 2 /n epi and 20 x 10 -11 cGy·cm 2 /n 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

  18. Evaluation of wrapper tube temperatures of fast neutron reactors using the TRANSCOEUR-2 code

    Energy Technology Data Exchange (ETDEWEB)

    Valentin, B.; Brun P. [CEA/DRN/DEC/SECA/LHC CEN, St Paul Lez Durance (France); Chaigne, G. [FRAMATOME/NOVATOME, Lyon (France)

    1995-09-01

    This paper deals with the thermal loading estimation of wrapper tubes using the TRANSCOEUR-2 code. This estimation requires a knowledge of two temperature fields: the first involves the peripheral sub-channel temperatures of each sub-assembly calculated by the design code CADET, and the second, outside the sub-assemblies, is the inter-wrapper flow temperature field calculated by the thermal-hydraulic code TRIO-VF with boundary conditions taken from CADET. Theoretical models of the three codes are presented as well as the first TRANSCOEUR-2 wrapper tube temperature calculation performed on the European Fast Reactor (EFR) Core Design 6/91 (CD 6/91) under nominal power conditions. The results show a temperature variation of 115{degrees}C between the bottom of the lower blanket and the top of the upper blanket fuel sub-assemblies in the center of the core and 95{degrees}C at the core periphery. The wrapper tube temperatures are higher in the center than in the external core.

  19. Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. To address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).

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

    International Nuclear Information System (INIS)

    Ibrahim, Mohammed

    2011-01-01

    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)

  1. Selectivity of benzene sulphonation in three gas—liquid reactors with different mass transfer characteristics: II: Mass transfer and selectivity in a cyclone reactor and in a tube reactor

    NARCIS (Netherlands)

    Beenackers, Antonie A.C.M.; van Swaaij, Willibrordus Petrus Maria

    1978-01-01

    Liquid benzene was sulphonated with gaseous sulphur trioxide in a tube reactor and in a new gas—liquid cyclone reactor. The products are benzenesulphonic acid and diphenyl sulphone (byproduct). The observed selectivity depends on the conversion, the initial benzene concentration and the mass

  2. An Inverse Fault Detection From Shallow Geophysical Data at the P Reactor Area, Savannah River Site, South Carolina

    Science.gov (United States)

    Cameron, A. E.; Knapp, C. C.; Addison, A. D.; Waddel, M.

    2008-12-01

    Surface and borehole Ground Penetrating Radar (GPR), a Shallow Seismic Reflection (SSR) as well as Electrical Resistivity Tomography (ERT) surveys were conducted at Savannah River Site (SRS), South Carolina, in order to investigate the shallow stratigraphy, hydrogeophysical zonation, and the applicability and performance of these imaging techniques. The study site is the P Reactor Area located within the Upper Atlantic Coastal Plain, with clastic sediments ranging from Late Cretaceous to Miocene age. Lithologies consist of sand, clayey sand, and clay with minor amounts of calcareous minerals. The target of this research is the delineation and prediction of migration pathways of a large contaminant plume including trichloroethylene (TCE) that originates from the northwest section of the reactor facility and discharges into the nearby Steel Creek. Here, we present results from stratigraphic and hydrogeophysical characterizations using (1) the GPR technique involving the PulseEKKO 100 GPR system with 50, 100, and 200 MHz antennas, (2) the SSR method with a Geometrics 120-channel seismograph, and (3) the ERT technique with the SuperSting R8 IP with 8-channel multi-electrode resistivity and IP imaging system. Faulting on the GPR, SSR and ERT profiles may be observed through offsets of reflecting interfaces, dipping discontinuities, or fault plane imaging, all leading to a constrained interpretation of a fault system at the study site. The simultaneous use of the 50, 100 and 200 MHz antennas with the SSR and ERT methods allows us to generate a geologic cross-section of the subsurface to perform analyes of the radar and acoustic reflection data as a function of frequency, conductivity and acoustic impedance to facilitate interpolation and extrapolation of the hydraulic properties such as hydraulic conductivity (K) and porosity of the study area.

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

    International Nuclear Information System (INIS)

    Ndibe, Collins; Maier, Jörg; Scheffknecht, Günter

    2015-01-01

    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 SO 2 mainly as a result of dilution. NO X 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 NO X emissions. In comparison to unstaged combustion, the nitrogen conversion to NO X 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 NO X 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

  4. Device and process for extracting a blocking sleeve of a nuclear reactor fuel assembly removable guide tube

    International Nuclear Information System (INIS)

    Petit, B.

    1990-01-01

    Removal the blocking sleeve of a nuclear reactor fuel assembly guide tube consits of deforming by bonding towards the inside of the blocking sleeve each of the zones of the formule having a deformed part projecting radially outwards in a manner to remove those deformed parts from location in the corresponding cavity of the tube plate. The blocking sleeve is then extracted by pulling it in the axial direction of the guide tube. A tod is used to bond it the sectors of the formule and grip them to allow extraction of the blocking sleeve by pulling. This extraction is performed under the water of a cooling pool [fr

  5. Thermal analysis of a helium-cooled, tube-bank blanket module for a tandem mirror fusion reactor

    International Nuclear Information System (INIS)

    Werner, R.W.

    1983-01-01

    A blanket module concept for the central cell of a tandem mirror reactor is described which takes advantage of the excellent heat transfer and low pressure drop characteristics of tube banks in cross-flow. The blanket employs solid Li 2 O as the tritium breeding material and helium as the coolant. The lithium oxide is contained in tubes arranged within the submodules as a two-pass, cross-flow heat exchanger. Primarily, the heat transfer and thermal-hydraulic aspects of the blanket design study are described in this paper. In particular, the analytical model used for selection of the best tube-bank design parameters is discussed in some detail

  6. Method and apparatus for testing closed-end tubes in heat exchangers of nuclear reactors and the like

    International Nuclear Information System (INIS)

    Seyd, G.; Bergbauer, A.; Paulsen, U.

    1975-01-01

    A description is given of a test stopper which is insertable into a tube closed at one end for testing the tightness of the tube with a fluid under pressure, the tube being in a heat exchanger of a nuclear reactor or the like. The test stopper includes a tubular outer jacket that is expandable outwardly to tightly seat the stopper in the tube. The stopper also has front and back end-face members joined to the ends of the outer jacket to define a closed space within the jacket. With the stopper inserted into the tube, the front end-face member and the closed end portion of the tube define a closed inner region of the tube. An inner tubular member, disposed within the outer jacket, partitions the closed space within the jacket into an annular outer chamber and a cylindrical inner chamber. A pressure-fluid supply selectively supplies fluid to the chambers. The outer jacket expands in response to fluid admitted to the annular chamber and the front end-face member has a through bore to admit fluid under pressure to the inner region of the tube. A method of testing of such a tube with a fluid under pressure includes inserting the test stopper into the tube and then expanding the outer jacket of the stopper to seat the stopper firmly in the tube. A fluid under pressure is directed through the stopper and into the closed region defined by the front end-face member of the stopper and the closed end portion of the tube. The pressure of the fluid introduced into this closed region is monitored for detecting a leak in the closed-end tube

  7. Reduction of pressure-tube to calandria-tube contact conductance to enhance the passive safety of a CANDU-PHW reactor

    International Nuclear Information System (INIS)

    Sanderson, D.B.; Moyer, R.G.; Litke, D.G.; Rosinger, H.E.; Girgis, S.

    1993-11-01

    One of the ways of enhancing the passive safety of a CANDU-PHW (Canada Deuterium Uranium-Pressurized heavy Water) reactor is to reduce the moderator subcooling requirements during a postulated loss-of-coolant accident (LOCA). The increased moderator temperatures would enhance the heat transfer from the moderator to the surrounding shield tank during a postulated accident. This reduction in subcooling requirements can be achieved by incorporating a wire screen in the fuel-channel annulus, right next to the calandria tube. This technique has been demonstrated to reduce significantly the moderator subcooling requirements, so that the calandria tube was not forced into film boiling upon pressure-tube ballooning contact with 0 degrees C subcooling outside the calandria tube. Two experiments, described in this report, were performed at AECL Research's Whiteshell Laboratories to investigate the changes in heat transfer characteristics between a pressure tube and a calandria tube, with a wire screen placed in the fuel-channel annulus. Results from computer simulations performed to assess the effect of the wire screen on the performance of a CANDU fuel channel during selected LOCA scenarios are also presented

  8. Steam-generator-tube-rupture transients for pressurized-water reactors

    International Nuclear Information System (INIS)

    Dobranich, D.; Henninger, R.J.; DeMuth, N.S.

    1982-01-01

    Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures

  9. Pressurized water reactor core instrument tube ruptures: Experimental simulation at the ROSA-IV LSTF

    International Nuclear Information System (INIS)

    Kukita, Y.; Asaka, H.; Nakamura, H.; Tasaka, K.

    1990-01-01

    A small-break loss-of-coolant accident (SBLOCA) initiated by the rupture of pressurized water reactor (PWR) core instrument tubes was simulated using the ROSA-IV Large Scale Test Facility (LSTF). The experimental results were characterized by a single-phase liquid discharge continuing for a long period until the pressure vessel coolant inventory was significantly depleted. This experiment was analyzed using three advanced LOCA analysis codes, RELAP5/MOD2, TRACPF1/MOD1 and CATHARE-1, to assess the predictive abilities of these codes. Although all these codes simulated the experimental results qualitatively, discrepancies were found between the predictions and data regarding the break flowrate, spatial distribution of coolant in the primary system and core rod temperature responses. (orig.)

  10. β-85 wt % Nb precipitates: the effect on in-reactor diametric creep of pressure tubes

    International Nuclear Information System (INIS)

    Sarce, Alicia

    2006-01-01

    By linking the microstructure evolution of an α-Zr crystal with the macroscopic behaviour, the deformation of an in-service reactor pressure tube is calculated. Microstructure evolution is considered through rate theory modelling of the interaction between point defects and sinks. Different densities of β-85 wt % Nb precipitates are proposed to be distributed inside the α grains and act as point defect sinks, doing a screening effect on the grain boundaries. From the interatomic pair potential which is used to describe the material, positive tangential deformation rates (on the other hand negatives) are obtained when these densities in the c-crystal direction are bigger than a minimum value. (author) [es

  11. 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, Jr., Thomas Dean [Univ. of Virginia, Charlottesville, VA (United States)

    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 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.

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

    International Nuclear Information System (INIS)

    Schulz, K.C.

    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 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

  13. Subcadmic and epicadmic flow in the dry tube of the TRIGA Mark III reactor of the Nuclear Center of Mexico

    International Nuclear Information System (INIS)

    Delfin L, A.; Mazon R, R.; Nava R, B.

    1991-04-01

    The mensuration of the thermal and fast flows of the irradiation facilities of the core of the reactor is important, since allow us to determine the optimum time of irradiation of the samples in the reactor. The Dry Tube especially, is an irradiation installation that it was designed in the I.N.I.N. to supply the pneumatic irradiation system of capsules with durations bigger than 15 minutes and it can be used for exposures until a maximum of three hours. The main users are the Nuclear Chemistry Department and the Neutron activation analysis. In this report the neutron flux sub cadmic and epi cadmic obtained in an experimental way in the Dry Tube for the reactor operating in stationary state to powers of 100 Kw, 300 Kw and 1000 Kw are reported and with these values it is interpolated for other powers. (Author)

  14. Savannah River reactor process water heat exchanger tube structural integrity margin Task Number 92-005-1

    International Nuclear Information System (INIS)

    Mertz, G.E.; Barnes, D.M.; Sindelar, R.L.

    1992-02-01

    Twelve process water heat exchangers are designed to remove heat generated in the reactor tank. Each heat exchanger has approximately 9000, 1/2 inch diameter x 0.049 inches thick tubes. Minimum structural tubing requirements and the leak rate through postulated tubing defects are developed in this report A comparison of the structural requirements and the defect size calculated to produce leak rates of 0.5 lbs./day demonstrate adequate structural margins against gross tube rupture. Commercial nuclear experience with pressurized water reactor (PWR) steam generator plugging criteria are used for guidance in performing this analysis. It is important to note that the SRS reactors are low energy systems with normal operating pressures of 203 psig at 130 degree F while the PWR is a high energy system with operating pressures near 2200 psig at 600 degree F. Clearly the PVM steam generator has loadings which are more severe than the SRS heat exchangers. Consistent with the Regulatory Guide 1.121 criteria both wastage (wall thinning) and cracking are addressed. Structural limits on wall thinning and crack size are developed to preclude gross rupture. ASME Section XI criteria, with the factors of safety recommended by Regulatory Guide 1.121 are used to develop the allowable crack size criteria. Normal operating conditions (pressure, dead weight, and hydraulic drag) are considered with seismic and water hammer accident conditions. Both the wall thinning and crack size criteria are developed for the end-of-evaluation period. Allowances for corrosion, wear, or crack growth have not been included in this analysis Structurally, the tubing is over designed and can tolerate large defects with adequate margins against gross rupture. The structural margins of heat exchanger tubing are evident by contrasting the tubing's structural capacity, per the ASME Code, with its operating conditions/configuration

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

    Energy Technology Data Exchange (ETDEWEB)

    Groenwall, B.; Ljungberg, L.; Huebner, W.; Stuart, W.

    1966-08-15

    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 {mu}g/cm{sup 2}). 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

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

    International Nuclear Information System (INIS)

    Groenwall, B.; Ljungberg, L.; Huebner, W.; Stuart, W.

    1966-08-01

    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/cm 2 ). 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

  17. Probe-holding apparatus for holding a probe for checking steam generator tubes particularly in a nuclear reactor installation

    International Nuclear Information System (INIS)

    Adamowski, A.; Gagny; Gallet, G.; Lhermitte, J.; Monne, M.; Vautherot, G.

    1984-01-01

    Probe-holding apparatus for holding a probe for checking steam generator tubes particularly in a nuclear reactor installation. The apparatus comprises a telescopic arm supported via a ball and socket joint from a support mounted in or near an access aperture in a chamber at one end of the steam generator. A probe guide is carried by a carriage pivotally mounted at the other end of the telescopic arm. The carriage includes an endless belt having a series of spaced projections which engage into the ends of the tubes, the projections being spaced by a distance equal to the tube pitch or a multiple thereof. The belt is driven by a stepping motor in order to move the carriage and place the probe guide opposite different ones of the tubes

  18. Design of an Irradiation Tube Assembly for the PTS no.1 and no. 2 at the HANARO Research Reactor

    International Nuclear Information System (INIS)

    Chung, Yong-Sam; Kim, Sun-Ha; Moon, Jong-Hwa; Jin, Byung-Jin; Kim, Hark-Rho

    2006-01-01

    Pneumatic transfer system (PTS) of a research reactor is one of the devices for a sample neutron irradiation. A newly designed PTS as one of the facilities to be used in an neutron irradiation of activation analysis has been developed for a reinstallation from the end of 2004. The design of a PTS is based on requirements such as the position of the irradiation hole and the geometry of the reactor, the neutron flux and distribution, a gamma heating and temperature at the irradiation place as well as the radiation dose rate, material and type of a rabbit, and the safety of the reactor operation, and so on. The basic composition consists of six systems as follows; 1) irradiation and transfer system (controller, irradiation tube, transfer tube, auto-loader, loader, receiver, air cushion valve assembly, diverter, photo sensor and a high purity polyethylene or graphite rabbit), 2) N2 gas supplier system, 3) gas exhaust system, 4) emergency system, 5) shielding system (loader-receiver, receiver, transfer line), 6) DNAA counting system. In this paper, the newly designed irradiation tube assembly (ITA) of PTS is presented and the results of a safety analysis in the HANARO research reactor are reviewed

  19. 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.

  20. Experimental investigation of the vibration response of a flexible tube due to simulated reactor core, cross and annular exit flows

    International Nuclear Information System (INIS)

    Haslinger, K.H.; Martin, M.L.; Higgins, W.H.; Rossano, F.V.

    1989-01-01

    Instrumentation tubes in pressurized nuclear reactors have experienced wear due to excessive flow-induced vibrations. Experiments to identify the predominant flow excitation mechanism at a particular plant, and to develop a sleeve design to remedy the wear problem are reported. An instrumented flow visualization model enabled simulation of a wide range of individual or combined reactor core flow, cross flow and thimble flow conditions. The instrumentation scheme adopted for these experiments used proximity displacement transducers and a force transducer to measure respectively tube motion and contact/impact forces at the wear region. Extensive testing of the original, in-plant configuration identified the normal core flow as the primary source of excitation. Shielding the In-Core-Instrumentation thimble tube from the normal core flow curtailed vibration amplitudes; however, thimble flow excitation then became more pronounced. Various outlet nozzle configurations were investigated. An internal cavity combined with radial outlet slots became the optimum solution for the problem. The paper presents typical test data in the form of orbital tube motion, spectrum analysis and time history collages. The effectiveness of shielding the instrumentation tube from the flow is demonstrated. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Shokralla, S., E-mail: shaddy.shokralla@opg.com [Ontario Power Generation, IMS NDE Projects, Ajax, Ontario (Canada); Krause, T.W., E-mail: thomas.krause@rmc.ca [Royal Military College of Canada, Kingston, Ontario (Canada)

    2014-01-15

    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)

  2. Control rod guide tube wear in operating reactors; operating experience report. Technical report December 1977-December 1979

    International Nuclear Information System (INIS)

    Riggs, R.

    1980-04-01

    Evidence of control rod guide tube wear has been observed in operating pressurized water reactors. The cause of this wear is identified as flow-induced vibration of the control rods. This report describes the measures being taken by both the industry and the NRC to deal with this matter. The staff also presents its technical positions and requirements to support continued operation of the plants as of December 1979 pending completion of this generic effort

  3. Cumulative damage fatigue tests on nuclear reactor Zircaloy-2 fuel tubes at room temperature and 3000C

    International Nuclear Information System (INIS)

    Pandarinathan, P.R.; Vasudevan, P.

    1980-01-01

    Cumulative damage fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature and 300 0 C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The cumulative cycle ratio at fracture for the Zircaloy-2 fuel tubes was found to depend on the sequence of loading, stress history, number of cycles of application of the pre-stress and the test temperature. A Hi-Lo type fatigue loading was found to be very much damaging at room temperature and this feature was not observed in the tests at 300 0 C. Results indicate significant differences in damage interaction and damage propagation under cumulative damage tests at room temperature and at 300 0 C. Block-loading fatigue tests are suggested as the best method to determine the life-time of Zircaloy-2 fuel tubes under random fatigue loading during their service in the reactor. (orig.)

  4. 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.

  5. Water-quality assessment of the Albemarle-Pamlico drainage basin, North Carolina and Virginia; trace elements in Asiatic clam (Corbicula fluminea) soft tissues and redbreast sunfish (Lepomis auritus) livers, 1992-93

    Science.gov (United States)

    Ruhl, P.M.; Smith, K.E.

    1996-01-01

    The analysis of potential contaminants in biological tissues is an important part of many water-quality assessment programs, including the National Water-Quality Assessment (NAWQA) Program. Tissue analyses often are used to provide information about (1) direct threats to ecosystem integrity, and (2) the occurrence and distribution of potential contaminants in the environment. During 1992-93, trace elements in Asiatic clam (Corbicula fluminea) soft tissues and redbreast sunfish (Lepomis auritus) livers were analyzed to obtain information about the occurrence and distribution of trace element contaminants in the Albemarle-Pamlico Drainage Basin of North Carolina and Virginia. The investigation was conducted as part of the NAWQA Program. All but 3 of the 22 trace elements that were analyzed were detected. Although all 10 of the U.S. Environmental Protection Agency (U.S. EPA) priority pollutants were detected in the tissues sampled, they were present in relatively low concentrations. Concentrations of U.S. EPA priority pollutants in Asiatic clams collected in the Albemarle-Pamlico Drainage Basin are similar to concentrations observed in other NAWQA study units in the southeastern United States. Mercury (a U.S. EPA priority pollutant) was widely detected, being present in 29 of 30 tissue samples, but concentrations did not exceed the FDA action level for mercury of a risk-based screening value for the general public. Mercury concentrations in Asiatic clams were similar to concentrations in other NAWQA study areas in the Southeast.

  6. Water-quality assessment of the Albemarle-Pamlico Drainage Basin, North Carolina and Virginia; organochlorine compounds in Asiatic clam (Corbicula fluminea) soft tissues and whole redbrest sunfish (Lepomis auritus) 1992-93

    Science.gov (United States)

    Smith, K.E.; Ruhl, P.M.

    1996-01-01

    The analysis of potential contaminants in biological tissues is an important part of many water-quality assessment programs, including the National Water-Quality Assessment (NAWQA) Program. Tissue analyses often are used to provide information about (1) direct threats to ecosystem integrity, and (2) the occurrence and distribution of potential contaminants in the environment. During 1992-93, Asiatic clam (Corbicula fluminea) soft tissues and whole redbreast sunfish (Lepomis auritus) samples were collected and analyzed to obtain information about the occurrence and distribution of organochlorine compounds in the Albemarle-Pamlico drainage Basin of North Carolina and Virginia. The investigation was conducted as part of the NAWQA Program. Relatively few organochlorine compounds were detected and of the compounds detected, all were detected in relatively low concentrations. The organochlorine compounds detected were p,p'-DDD, p,p'-DDE, p,p'-DDT, dieldrin, trans-nonachlor, PCB's, and toxaphene. Multiple compounds were detected at 16 of 19 sites sampled. Compared to Asiatic clams, redbreast sunfish appear to be better bioindicators of organochlorine contamination in aquatic systems. Except for one detection of toxaphene, pesticide concentrations are well below the National Academy of Sciences and National Academy of Engineering (NAS/NAE) guidelines for the protection of fish-eating wildlife.

  7. Process Intensification of Alkynol Semihydrogenation in a Tube Reactor Coated with a Pd/ZnO Catalyst

    Directory of Open Access Journals (Sweden)

    Nikolay Cherkasov

    2017-11-01

    Full Text Available Semihydrogenation of 2-methyl-3-butyn-2-ol (MBY was studied in a 5 m tube reactor wall-coated with a 5 wt% Pd/ZnO catalyst. The system allowed for the excellent selectivity towards the intermediate alkene of 97.8 ± 0.2% at an ambient H2 pressure and a MBY conversion below 90%. The maximum alkene yield reached 94.6% under solvent-free conditions and 96.0% in a 30 vol % MBY aqueous solution. The reactor stability was studied for 80 h on stream with a deactivation rate of only 0.07% per hour. Such a low deactivation rate provides a continuous operation of one month with only a two-fold decrease in catalyst activity and a metal leaching below 1 parts per billion (ppb. The excellent turn-over numbers (TON of above 105 illustrates a very efficient utilisation of the noble metal inside catalyst-coated tube reactors. When compared to batch operation at 70 °C, the reaction rate in flow reactor can be increased by eight times at a higher reaction temperature, keeping the same product decomposition of about 1% in both cases.

  8. Process to remote the guide pins of a guide tube being part of the upper internal equipments of a PWR reactor and the associated device

    International Nuclear Information System (INIS)

    Styskal, P.; Guicherd, L.; Clar, G.

    1984-01-01

    The upper internal equipments are set above the reactor core. They consist of long vertical guide tubes to guide the control rods located above a specific fuel element cluster of the core. Each tube consists of a lower and an upper tube length, the lower length being fixed by the guide pins in the top reactor core plate, while the upper length, constituting an extension of the lower one, rests on a horizontal support plate attached by a bracing structure to the upper core plate. The upper and lower lengths of the tubes are connected independently of one other, to the support plate by bolts which also connects these two parts together. During shutdown, the entire upper internal equipments are removed from the pressure vessel, the reactor pool and the vessel being filled with water, and placed at a storage post in the pool of the nuclear reactor. A temporary tube locking assembly, consisting of a vertical rod and end clamping pieces is inserted into the two tube lengths, thus locking them together, whereupon the bolts are removed and the connection between tube lengths and plate is thus released. The tube is now raised slightly, thus disconnecting its bottom guide pins from the reactor core plate and the entire tube is transferred by crane to a different zone, where it is decontaminated, dried and fitted with new guide pins at the bottom. The tube is then replaced underwater in the maintenance bay, the bolts are inserted again and the temporary structure is removed. The invention applies to PWR reactor maintenance, during shutdown for refueling [fr

  9. Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Office of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high

  10. Virginia Tech To Showcase Northern Virginia Research

    OpenAIRE

    Cunningham, Deborah

    2003-01-01

    Virginia Tech graduate students and faculty members in Northern Virginia will display and explain their most current research at the first Virginia Tech at Northern Virginia Research Exposition April 17 at the Northern Virginia Center in Fairfax.

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

    International Nuclear Information System (INIS)

    Flach, G.P.

    1999-01-01

    A regional groundwater flow model encompassing approximately 100 mi 2 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Flach, G.P.

    1999-02-24

    A regional groundwater flow model encompassing approximately 100 mi{sup 2} 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.

  13. Record of Decision; Continued operation of K, L, and P Reactors, Savannah River Site, Aiken, South Carolina

    International Nuclear Information System (INIS)

    1991-01-01

    The US Department of Energy (DOE) has considered the environmental impacts, benefits and costs, and institutional and programmatic needs associated with continued operation of the Savannah River Site (SRS) reactors, and has decided that it will continue to operate K and L Reactors at SRS, and will terminate operation of P Reactor in the immediate future and maintain it in cold standby. For P Reactor, this will involve the reactor's defueling; storage of its heavy water moderator in tanks in the reactor building; shutdown of reactor equipment and systems in a protected condition to prevent deterioration; and maintenance of the reactor in a defueled, protected status by a skeleton staff, which would permit any future decision to refuel and restart. Currently committed and planned upgrade activities will be discontinued for P Reactor. DOE will proceed with the safety upgrades and management system improvements currently scheduled for K Reactor in its program to satisfy the criteria of the Safety Evaluation Report (SER), and will conduct an Operational Readiness Review (ORR). The satisfaction of the SER criteria and completion of the ORR will demonstrate that the safety and health criteria for the resumption of production have been met. Reactor restart is expected to be in the third quarter of 1991 for K Reactor

  14. 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)

    Takeda, Takeshi; Furusawa, Takayuki

    2001-07-01

    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)

  15. Design and Computational Fluid Dynamics Optimization of the Tube End Effector for Reactor Pressure Vessel Head Type VVER-1000

    International Nuclear Information System (INIS)

    Novosel, D.

    2006-01-01

    In this paper is presented development and optimization of the tube end effector design which should consist of 4 ultrasonic transducers, 4 Eddy Current's transducers and Radiation Proof Dot Camera. Basically, designing was conducted by main input requests, such as: inner diameter of a tested reactor pressure vessel head penetration tube, dimensions of a transducers and maximum allowable vertical movement of a manipulator connection rod in order to cover all inner tube surface. As is obvious, for ultrasonic testing should be provided the thin layer of liquid material (in our case water was chosen) which is necessary to make physical contact between transducer surface and investigated inner tube surface. By help of Computational Fluid Dynamics, determined were parameters of geometry, as the most important factor of transducer housing, hydraulically parameters for water supply and primary drain together implemented into this housing, movement of the end effectors (vertical and cylindrical) and finally, necessary equipment which has to provide all hydraulically and pneumatic requirements. As the cylindrical surface of the inner tube diameter was liquefied and contact between transducer housing and tested tube wasn't ideally covered, water leakage could occur in downstream direction. To reduce water leakage, which is highly contaminated, developed was second water drain by diffuser assembly which is driven by Venturi pipe, commercially called vacuum generator. Using the Computational Fluid Dynamic, obtained was optimized geometry of diffuser control volume with the highest efficiency, in other words, unobstructed fluid flux. Afterwards, the end effectors system was synchronized to the existing operable system for NDT methods all invented and designed by INETEC. (author)

  16. Proposals for investigating instrument tube line breaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Charlton, T.R.; Loomis, G.G.; Hall, D.G.; Cozzuol, J.M.

    1985-11-01

    Questions posed by the NRC pertaining to instrument tube critical flow and applicability of the Semiscale experimental facility are evaluated. A program is recommended to investigate the issue of generic PWR safety following hypothetical rupture of instrument tubes due to consequences of seismic events

  17. A flow-through amperometric sensor based on dialysis tubing and free enzyme reactors

    NARCIS (Netherlands)

    Bohm, S.; Pijanowska, D.G.; Pijanowska, D.; Olthuis, Wouter; Bergveld, Piet

    2001-01-01

    A generic flow-through amperometric microenzyme sensor is described, which is based on semi-permeable dialysis tubing carrying the sample to be analyzed. This tubing (300 μm OD) is led through a small cavity, containing the working and reference electrode. By filling this cavity with a few μl of an

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

    International Nuclear Information System (INIS)

    Homonnay, Z.; Kuzmann, E.; Varga, K.; Nemeth, Z.; Szabo, A.; Rado, K.; Schunk, J.; Tilky, P.; Patek, G.

    2005-01-01

    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. 57 Fe-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.

  19. 75 FR 15704 - Old Dominion Electric Cooperative; North Carolina Electric Membership Corporation, Complainants v...

    Science.gov (United States)

    2010-03-30

    ... Electric Cooperative; North Carolina Electric Membership Corporation, Complainants v. Virginia Electric and... the Federal Power Act, 16 U.S.C. 824(e) and 825(e), Old Dominion Electric Cooperative and North Carolina Electric Membership Corporation (Complainants) filed a formal complaint against Virginia Electric...

  20. Leak-before-break concept for evaluation of flows detected in pressure tubes in a Candu type reactor

    International Nuclear Information System (INIS)

    Crespi, J.C.

    1992-01-01

    This paper reviews the role of the Leak-Before-Break concept for evaluation of flaws detected in cold-worked Zr 2.5% Nb pressure tubes in a CANDU type reactors. The acceptance criteria are intended to prevent failure by brittle fracture, plastic collapse of the ligament and delayed hydride cracking. The methodology developed here was of great help in order to establish the operative conditions for fuel channel garter springs repositioning by means of the SLA Rette tool at Embalse Nuclear Generating Station, Cordoba, Argentina. (author)

  1. Visual inspection technology of the narrow and small confined area for monitoring feederpipe support of pressure tube in calandria reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Wan; Lee, Nam Ho; Choi, Young Soo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    There are 760 feederpipes, which they are connected to inlet/outlet of the 380 pressure tube channels on the front of the calandria, in CANDU-type Reactor of Wolsung Nuclear Power Plant. As an ISI(In-Service Inspection) and PSI (Post-Service Inspection) requirements, maintenance activities of measuring the thickness of curvilinear part of feederpipe and inspecting the feederpipe support area within calandria are needed to ensure continued reliable operation of nuclear power plant. And ultrasonic probe is used to measure the thickness of curvilinear part of feederpipe, however workers are exposed to radioactivity irradiation during the measurement period. But, it is exposed to radioactivity irradiation during the measurement period. But, it is impossible to inspect feederpipe support area thoroughly because of narrow and confined accessibility, that is , an inspection space between the pressure tube channels is less than 100 mm and pipes in feederpipe support area are congested. And also, workers involved in inspecting feederpipe support area are under the jeopardy of high-level radiation exposure. Concerns about sliding home, which make the move of feederpipe connected to pressure tube channel smooth as pressure tube expands and contracts in its axial direction, stuck to feedeerpipe support and some of the structural components have made necessary the development of video inspection probe system with narrow and confined accessibility to observe and inspect feederpipe support area more close. Using video inspection probe system, it is possible to inspect and repair abnormality of feederpipe support connected to pressure tube channels of the calandria more accurate and quantative than naked eye. Therefore, that will do much for ensuring safety of CANDU-type nuclear power plant. 45 figs.,31 tabs. (Author)

  2. Friction pressure drop and heat transfer coefficient of two-phase flow in helically coiled tube once-through steam generator for integrated type marine water reactor

    International Nuclear Information System (INIS)

    Nariai, Hideki; Kobayashi, Michiyuki; Matsuoka, Takeshi.

    1982-01-01

    Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report. (author)

  3. Co-pyrolysis behaviors of saw dust and Shenfu coal in drop tube furnace and fixed bed reactor.

    Science.gov (United States)

    Li, Shuaidan; Chen, Xueli; Wang, Li; Liu, Aibin; Yu, Guangsuo

    2013-11-01

    Co-pyrolysis behaviors of saw dust (SD) and Shenfu bituminous coal (SF) were studied in a drop tube furnace and a fixed bed reactor at different temperatures respectively. Six different biomass/coal ratios (B:C) were used. Compared the results with the calculated value obtained by the additional behavior, CO volume yields were lower while H2, CH4, CO2, volume yields were higher. Blend char yields had a good agreement with the calculated values, and their structures remained similar with SD and SF char's. Synergy effect occurred in gaseous phase, which was mainly caused by the secondary reactions. Compared the blend char yields in the drop tube furnace with those in the fixed bed reactor, the results showed the contacting way of biomass and coal particles had little influence on char yield in co-pyrolysis process. The reactivity index of blend char achieved the minimum at B:C=40:60 and the maximum at B:C=80:20. Copyright © 2013 Elsevier Ltd. All rights reserved.

  4. Water-quality assessment of the Albermarle-Pamlico drainage basin, North Carolina and Virginia; environmental setting and water-quality issues

    Science.gov (United States)

    McMahon, Gerard; Lloyd, Orville B.

    1995-01-01

    -sediment yields for selected forested, agricultural, and developed urban basins in North Carolina are 50, 250, and 550 tons per square mile, respectively. In order to facilitate comparisons, much of the data were compiled by hydrologic unit. Homogeneous areas, or strata, representing the most prevalent combinations of environmental factors, such as land use, soils, and geology, were defined. Future data collection and analyses will be designed to answer objective-related concerns about the relations between important water-quality conditions and these study-unit strata.

  5. 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

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-02-01

    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.

  6. 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)

    Monteleone, S.

    1997-02-01

    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

  7. 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.

  8. In reactor measurements, modeling and assessments to predict liquid injection shutdown system nozzle to Calandria tube time to contact

    International Nuclear Information System (INIS)

    Kirstein, K.; Kalenchuk, D.

    2011-01-01

    Over the past few years there has been an expanding effort to assess the potential for Calandria Tubes (CTs) coming into contact with Liquid Injection Shutdown System (LISS) Nozzles to ensure continued contact-free operation as required by CSA N285.4. LISS Nozzles (LINs), which run perpendicular to and between rows of fuel channels, sag at a slower rate than the fuel channels. As a result certain LINs may come in contact with CTs above them. The CT/LIN gaps can be predicted from calculated CT sag, LIN sag and a number of component and installation tolerances. This method however results in very conservative predictions when compared to measurements, confirmed with the in reactor measurements initiated in 2000, when gaps were successfully measured the first time using images obtained from a camera-assisted measurement tool inserted into the calandria. To reduce the conservatism of the CT/LIN gap predictions, statistical CT/LIN gap models are used instead. They are derived from a comparison between calculated gaps based on nominal dimensions and the visual image based measured gaps. These reactor specific (typically 95% confidence level) CT/LIN gap models account for all uncertainties and deviations from nominal values. Prediction error margins reduce as more in-reactor gap measurements become available. Each year more measurements are being made using this standardized visual CT/LIN proximity method. The subsequently prepared reactor-specific models have been used to provide time to contact for every channel above the LINs at these stations. In a number of cases it has been used to demonstrate that the reactor can be operated to its end of life before refurbishment with no predicted contact, or specific at-risk channels have been identified for which appropriate remedial actions could be implemented in a planned manner. (author)

  9. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  10. R Reactor seepage basins soil moisture and resistivity field investigation using cone penetrometer technology, Savannah River Site, Aiken, South Carolina

    International Nuclear Information System (INIS)

    Harris, M.K.

    2000-01-01

    The focus of this report is the summer 1999 investigation of the shallow groundwater system using cone penetrometer technology characterization methods to determine if the water table is perched beneath the R Reactor Seepage Basins (RRSBs)

  11. R Reactor seepage basins soil moisture and resistivity field investigation using cone penetrometer technology, Savannah River Site, Aiken, South Carolina

    Energy Technology Data Exchange (ETDEWEB)

    Harris, M.K.

    2000-02-17

    The focus of this report is the summer 1999 investigation of the shallow groundwater system using cone penetrometer technology characterization methods to determine if the water table is perched beneath the R Reactor Seepage Basins (RRSBs).

  12. Metallurgical problems in the exchange tube of a fast reactor steam generator

    International Nuclear Information System (INIS)

    Coriou, M.; Champeix, L.; Weisz, M.

    1980-10-01

    The design of the 1200 MWe Super Phenix power station steam generators is based on the following principles: once through helical tube exchangers which can be completely drained on the sodium side; the single wall exchange tubes are accessible to Foucault current testing during shutdowns. The authors explain the reasons for selecting the 800 Alloy for the exchange tubes. This choice was borne out by the results of several years of studies in the following areas: 6000 test hours with a 45 MWe model; corrosion test under stress in a water-steam and sodium plus caustic soda environment; resistance to creep and fatigue (effects of ageing and annealing, of the chemical compound); industrial feasibility, fabrication, utilization, bending, coiling, welding, testing. Concurrently, the EMl2 qualification finalizing has been pursued for the same application [fr

  13. Testing external surface of fuel element tubes for power nuclear reactors

    International Nuclear Information System (INIS)

    Naugol'nykh, O.G.; Nelyubin, Yu.V.

    1987-01-01

    Optical methods are regarded perspective for discovery and detection of flaws of external surfaces of fuel element tubes. The TV method has highest information content among them. Two mock-ups of facilities based on the TV method using a ''dissector'' type TV device and a TV tube with charge accumulation (vidikon) have been developed. It is concluded that complex testing - combination of ultrasonic, photoelectric and TV methods in a facility is necessary for discovery and analysis of the whole variety of flaws, though sensitivity of the TV method is enough for disclosure of all the main defects

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

    International Nuclear Information System (INIS)

    Prot, A.; Monnier, P.

    1964-01-01

    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) [fr

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

    International Nuclear Information System (INIS)

    2009-03-01

    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'

  16. Preliminary design study of removable integral steam generator units of the multiple helically wound tube type for a 1250 MW(th) H.T.G.C. reactor

    International Nuclear Information System (INIS)

    Gilli, P.V.; Fritz, K.; Lippitsch, J.; Sandri, A.H.; Weiss, B.

    1965-11-01

    The possibilities of designing a multiple steam generator for a 1250 MW(th) High Temperature Gas-Cooled Reactor, consisting of 18 units which are able to pass through 5 ft diam. holes in the integral prestressed concrete pressure vessel are investigated. A lay-out and design with bundles of multi-start helical tubes is evolved, particular attention being paid to the questions of tube blanking and removal of the unit, and of selection of materials for superheater and reheater tubes. Thermal and stress calculations have been carried out, using the Waagner-Biro Computer Code ADURHELIX. (author)

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

    International Nuclear Information System (INIS)

    1994-01-01

    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

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

    International Nuclear Information System (INIS)

    Spriet, M.

    1975-01-01

    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) [fr

  19. A flow-through amperometric sensor based on dialysis tubing and free enzyme reactors.

    Science.gov (United States)

    Böhm, S; Pijanowska, D; Olthuis, W; Bergveld, P

    2001-08-01

    A generic flow-through amperometric microenzyme sensor is described, which is based on semi-permeable dialysis tubing carrying the sample to be analyzed. This tubing (300 microm OD) is led through a small cavity, containing the working and reference electrode. By filling this cavity with a few microl of an appropriate enzyme solution, an amperometric enzyme sensor results. As the dialysis tubing is impermeable for large molecular species such as enzymes, this approach does not require any immobilization chemistry, and as a consequence the enzyme is present in its natural free form. Based on this principle, amperometric sensors for lactate, glucose, and glutamate were formed by filling cavities, precision machined in Perspex, with buffered solutions containing respectively, lactate-, glucose-, and glutamate-oxidase. All sensors showed a large linear range (0-35 mM for glucose, 0-3 mM for lactate, and 0-5 mM for glutamate) covering the complete physiological range. The lower detection limit was in the order of 15-50 microM. Applicability in flow injection analysis systems is demonstrated.

  20. Steam generator assembly for pressurized water reactors with a straight tube bundle and a partial flow preheater traversible by pressurized water

    International Nuclear Information System (INIS)

    Michel, R.

    1976-01-01

    To reduce the temperature difference between a straight tube bundle and the housing surrounding the same in a steam generator assembly for pressurized water reactors, a preheater for feed water is provided, and part of the pressurized water, after it has flowed through the heat exchanger or steam generator proper, is used for heating the feedwater in the preheater. 3 claims, 1 drawing figure

  1. Effects of Relative SG Tube Pitches on the Performance Characteristics of a Small Modular Reactor driven by Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Youngjin; Yi, Kunwoo; Lee, Byungjin [KEPCO Engineering and Construction, Daejeon (Korea, Republic of)

    2016-10-15

    In this research, the capacity and basic dimensions for SMRs driven by a natural circulation are preliminarily assumed to determine the SMR configuration for the conceptual design, and each of the pre-set values is explained below. Firstly, the PZR configuration is not considered because it is not included to the main flow of the primary coolant. One of the SMR requirements is that SMR shall carry on the road. Hence, the vehicle geometrical limits are 15 meters for the length, and 3.5 meters for the height, approximately. With these limits for the dimensions of the SMR, RV length is assumed about 13.8 meters and RV diameter about 2.5 meters. In IAEA definition for SMRs, the capacity of electric power is no more than 300 MWe. If the efficiency of SMR power plant is assumed to 33% compared to the commercial power plant, the core power is below 1,000 MWth. In this research, the core power is assumed to 200 MWth arbitrarily during normal operation. The primary coolant passes through the outside of tubes, and the heat is transfer to the secondary feedwater. The secondary feedwater passes through the inside of tubes, and the heat from the primary coolant is received to generate the superheated steam. The present work carries out numerical simulations to get an insight for the effects of the diameters of the reactor vessel and riser using the parameters such as the steam generator tube pitches. To sum up, the calculation results show a good agreement with the theoretical equation and the uniform diameter loop has a more uniform temperature distribution and larger mass flow rate.

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

    International Nuclear Information System (INIS)

    Subbotin, V.; Alexeev, G.; Peskov, O.; Sapankevic, A.

    1976-01-01

    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.)

  3. A calorimeter for determination of heating of materials in the irradiation tubes of the reactor FR 2

    International Nuclear Information System (INIS)

    Kapulla, H.; Schuelken, H.

    1965-01-01

    For technical and economic reasons it is desirable to know the rate of absorption of energy by different materials in different places in a reactor. It is often hard to calculate this rate because data on the flux and spectra of the radiation are not sufficiently well known. However, the rate can be measured with a calorimeter. By choosing different absorbing materials, some success can be achieved in determining the energy contribution of the different kinds of radiation. Thus, for example, bismuth and lead have high absorption coefficients for γ-rays, the scattering of fast neutrons can be ignored and their activation cross-sections are low, so they are suitable for measuring the energy contribution of the γ-rays. On the other hand, for neutron dosimetry the lighter elements, such as beryllium or hydrogen. must be used. In this case, however, the effect of γ-rays cannot be ignored, which makes interpretation of the results more difficult. For the measurements of the heating of materials in the Irradiation tubes of the FR-2 reactor an isothermal type of calorimeter was chosen. The mechanical construction of the calorimeter is shown. Bismuth, iron, aluminium, and graphite are used as calorimeter materials. The design is such that the rods can be changed easily. The temperatures of the aluminium can, and the calorimeter foot, are regulated independently by electrical heating and forced draft cooling, and can be held at 50±0.02 deg C. The calibration of the calorimeter was achieved by heating the rod electrically, turning off the current, and observing the exponential cooling rate from which the cooling time constant could be determined. The aluminium can must be kept at the same temperature during calibration as during measurements in the reactor. Temperatures were measured with iron constantan thermocouples. Measurements were made in some of the channels of the FR-2 using this calorimeter. The measured energy absorption rates at the core centre are between

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

    International Nuclear Information System (INIS)

    2006-05-01

    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

  5. Statistical analysis of failure time in stress corrosion cracking of fuel tube in light water reactor

    International Nuclear Information System (INIS)

    Hirao, Keiichi; Yamane, Toshimi; Minamino, Yoritoshi

    1991-01-01

    This report is to show how the life due to stress corrosion cracking breakdown of fuel cladding tubes is evaluated by applying the statistical techniques to that examined by a few testing methods. The statistical distribution of the limiting values of constant load stress corrosion cracking life, the statistical analysis by making the probabilistic interpretation of constant load stress corrosion cracking life, and the statistical analysis of stress corrosion cracking life by the slow strain rate test (SSRT) method are described. (K.I.)

  6. Macromixing hydrodynamic study in draft-tube airlift reactors using electrical resistance tomography.

    Science.gov (United States)

    Gumery, Farouza; Ein-Mozaffari, Farhad; Dahman, Yaser

    2011-02-01

    The present study summarizes results of mixing characteristics in a draft tube airlift bioreactor using ERT. This technique offers the possibility for noninvasive and nonintrusive visualization of flow fields in the bioreactor and has rarely been utilized previously to analyze operating parameters and mixing characteristics in this type of bioreactors. Several operating parameters and geometric characteristics were examined. In general, results showed that the increase in superficial gas velocity corresponds to an increase in energy applied and thus, to a decrease in mixing time. This generally corresponded to an increase in liquid circulation velocity and shear rate values. Bottom clearances and draft tube diameters affected flow resistance and frictional losses. The influence of sparger configurations on mixing time and liquid circulation velocity was significant due to their effect on gas distribution. However, the effect of sparger configuration on shear rate was not significant, with 20% reduction in shear rates using the cross-shaped sparger. Fluid viscosity showed a marked influence on both mixing times and circulation velocity especially in the coalescing media of sugar and xanthan gum (XG) solutions. Results from this work will help to develop a clear pattern for operation and mixing that can help to improve several industrial processes, especially the ones related to emerging fields of technology such as the biotechnology industry.

  7. Numerical study on pressure drop and heat transfer for designing sodium-to-air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kang, Hie-Chan; Eoh, Jae-Hyuk; Cha, Jae-Eun; Kim, Seong-O.

    2013-01-01

    Highlights: ► Numerical simulation for the heat flow characteristic of the sodium-to-air heat exchanger (AHX) and tube banks. ► Parallelogram tube banks showed almost similar thermal and hydraulic characteristics to the rectangular tube banks. ► Pressure drop and heat transfer of the staggered and rectangular tube banks compared with Zhukauskas’ correlation. ► AHX was modeled as porous media and suggested design guide to enhance the performance. - Abstract: A numerical study is 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 are modeled as porous media and simulated heat and momentum transfer by a commercial program. Two-dimensional flow characteristic appears differently at the inlet region of the AHX annulus, and the required length of the inlet region is shorter for an inlet having a 45 degree chamber or a round shape than for one with a perpendicular corner. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX are evaluated and discussed. Pressure drop and heat transfer shows similar trends and underestimated values, respectively, when compared with Zhukauskas empirical correlations. The parallelogram tube bank shows similar results to the rectangular arrangement.

  8. High pressure thimble/guide tube seal fitting with built-in low pressure seal especially suitable for facilitated and more efficient nuclear reactor refueling service

    International Nuclear Information System (INIS)

    Bhatt, P.N.; Blaushield, R.M.

    1991-01-01

    This patent describes a HP/LP seal arrangement for an elongated guide tube and an elongated thimble disposed therein. The guide tube and thimble extending outwardly from the core of a nuclear reactor to a seal table where the guide tube is welded to the seal table to provide a high pressure seal relative thereto. It comprises: a tubular seal fitting disposed in alignment with the guide tube with the thimble extending therethrough on the low pressure side of the seal table; first high pressure sealing means coupling one end of the fitting to an end of the guide tube to prevent leakage from within the guide tube; inwardly facing thread means disposed adjacent the other and outer end of the seal fitting; a nut having an opening through which the thimble extends and further having outwardly facing threading in mating engagement with the fitting thread means; the fitting having a seal seat spaced longitudinally inwardly from the thread means and facing the fitting outer end and further disposed annularly about the inner surface of the fitting; deformable ring seal means; second releasable high pressure sealing means coupling the thimble to the outer end portion of the guide tube

  9. Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type boiling water reactor

    International Nuclear Information System (INIS)

    Jain, Vikas; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2011-01-01

    Highlights: → We have highlighted the problem of drum level dynamics in a multiple loop type NC system using RELAP5 code. → The need of interconnections in steam and liquid spaces close to drum is established. → The steam space interconnections equalize pressure and liquid space interconnections equalize level. → With this scheme, the system can withstand anomalous conditions. → However, the controller is found to be inevitable for inventory balance. - Abstract: Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main heat transport system (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam-water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam-water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.

  10. Optimization of the Manufacturing Process of Zr-2.5Nb Pressure Tubes for CANDU Reactors for Extending Their Design Life to Over 30 years

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-10-15

    Zr-2.5Nb pressure tubes are the most critical components that determine the design life of CANADU (CAnadian Natural Uranium) reactors. The initial design target for the Zr-2.5Nb pressure tubes is to suppress the diametral creep through a texture control which may trade off the other performances that can be overcome by introducing a change in the components design. To this end, they are made by the extrusion process at high temperatures to have a circumferential texture with most of the basal poles oriented towards their circumferential direction. However, this circumferential texture causes them to be very susceptible to delayed hydride cracking (DHC) and to have a higher axial elongation. Against the initial design target, their costly refurbishments are planned in several commercial CANDU reactors before their design life of 30 years, due to the unexpectedly faster creep rate and axial elongation. This fact casts a question over the validity of the design philosophy that the diametral creep of the Zr-2.5Nb pressure tube is governed by the texture. The aim of this work is to elucidate the governing factor of creep of the Zr-2.5Nb tubes and to find a way of making improved Zr-2.5Nb pressure tubes with a lower diametral creep and axial elongation. To this end, we scrutinized Holt's experiment where the in-reactor creep behaviors of the Zr-2.5Nb micro-pressure tube (MPT) with a circumferential texture was compared with that of the Zr-2.5Nb fuel sheath (FS) with a radial texture. Accounting for the fact that thermal creep of Zr-2.5Nb alloy is affected by the Nb concentration in the {beta}-Zr, we demonstrate that the reduced creep is not dictated by the circumferential texture but by the increased Nb concentration in the {alpha}-Zr. This study suggests that the optimized manufacturing procedure of the Zr-2.5Nb tube would improve their in-reactor performances, extending their design life to over 30 years when compared to that of the current design of the

  11. Current status of materials development of nuclear fuel cladding tubes for light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Zhengang, E-mail: duan_zg@imr.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Yang, Huilong [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Satoh, Yuhki [Institute for Materials Research, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Abe, Hiroaki, E-mail: abe.hiroaki@n.t.u-tokyo.ac.jp [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan)

    2017-05-15

    Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

  12. Development of Bundle Position-Wise Linear Model for Predicting the Pressure Tube Diametral Creep in CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Yong [Korea Electric Power Corporation Research Institute, Daejeon (Korea, Republic of); Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

    2011-08-15

    Diametral creep of the pressure tube (PT) is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of a heat transport system. PT diametral creep leads to diametral expansion that affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux. Therefore, it is essential to predict the PT diametral creep in CANDU reactors, which is caused mainly by fast neutron irradiation, reactor coolant temperature and so forth. The currently used PT diametral creep prediction model considers the complex interactions between the effects of temperature and fast neutron flux on the deformation of PT zirconium alloys. The model assumes that long-term steady-state deformation consists of separable, additive components from thermal creep, irradiation creep and irradiation growth. This is a mechanistic model based on measured data. However, this model has high prediction uncertainty. Recently, a statistical error modeling method was developed using plant inspection data from the Bruce B CANDU reactor. The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. There are twelve bundles in a fuel channel and for each bundle, a linear model was developed by using the dependent variables, such as the fast neutron fluxes and the bundle temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3 and 4 were used to develop the BPLM models. The remaining 10 channels' data were used to test the developed BPLM models. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from the Units 2,3 and 4 in Korea. Two error components for the BPLM, which are the

  13. 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.

  14. Reactors

    International Nuclear Information System (INIS)

    Onuki, Koji; Sasanuma, Katsumi.

    1980-01-01

    Purpose: To make it possible to correctly measure the flow rate and temperatures of the coolants flowing through fuel assemblies. Constitution: One or more holes are formed at the side surface of the guide tube of a control rod driving mechanism thereby to reduce the flow path resistance within the guide tube of the control rod driving mechanism and to prevent the outlet coolant of the control rod guide tube from flowing into the guide tube of the mechanism as it is and also from flowing into ambient rectifying lattice guide tubes, so that the quantities and temperatures of the coolants flowing through respective fuel assemblies can be measured correctly. (Kamimura, M.)

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

    International Nuclear Information System (INIS)

    Takizuka, T.; Tokunaga, S.; Hoshino, K.; Shimizu, K.; Asakura, N.

    2015-01-01

    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

  16. Radiation damage studies of straw tube and scintillating fiber elements

    International Nuclear Information System (INIS)

    Dunn, W.L.; Elleman, T.S.; Goshaw, A.T.; Oh, S.H.; Robertson, W.J.; Grimes, A.; Leedom, I.; Reucroft, S.

    1990-01-01

    The authors report on the results of mixed-field irradiations of straw-tube, plastic scintillating fiber, and avalanche photodiode components. These irradiations are being carried out at the one-MW PULSTAR research reactor facility at North Carolina State University. A special sample holder was designed that allows relatively uniform irradiation of samples up to 5 ft long, without bending or coiling. A systematic irradiation program is underway that allows study of total fluence, fluence-rate, and neutron spectral effects. Samples have been exposed to neutron fluences as high as 2 x 10 16 cm -2

  17. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  18. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  19. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1984-10-01

    A review of the performance of steam generator tubes in 116 water-cooled nuclear power reactors showed that tubes were plugged at 54 (46 percent) of the reactors. The number of tubes removed from service decreased from 4 692 (0.30 percent) in 1981 to 3 222 (0.20 percent) in 1982. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that have used only volatile treatment, with or without condensate demineralization

  20. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1983-08-01

    A review of the performance of steam generator tubes in 110 water-cooled nuclear power reactors showed that tubes were plugged at 46 (42 percent) of the reactors. The number of tubes removed from service increased from 1900 (0.14 percent) in 1980 to 4692 (0.30 percent) in 1981. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that used all-volatile treatment since start-up. At one reactor a large number of degraded tubes were repaired by sleeving which is expected to become an important method of tube repair in the future

  1. Fermentative hydrogen production with a draft tube fluidized bed reactor containing silicone-gel-immobilized anaerobic sludge

    Energy Technology Data Exchange (ETDEWEB)

    Lin, Chi-Neng; Wu, Shu-Yii [Department of Chemical Engineering, Feng Chia University, Taichung (China); Chang, Jo-Shu [Department of Chemical Engineering, National Cheng Kung University, Tainan (China)

    2006-12-15

    A draft tube fluidized bed reactor (DTFBR) containing immobilized cell particles was designed to produce H{sub 2} continuously. A synthetic polymer (silicone gel; SC) was used as the primary material to immobilize acclimated anaerobic sludge for H{sub 2} production in DTFBR with a working volume of 8L. The DTFBR system was operated at a hydraulic retention time (HRT) of 2.2-8.9h and an influent sucrose concentration (C{sub s}) of 5-40g COD/l. The results show that in general decreasing HRT or increasing sucrose concentration led to a marked increase in the volumetric H{sub 2} production rate (v{sub H{sub 2}}), but a gradual decrease in the H{sub 2} yield (Y{sub H{sub 2}}). The best v{sub H{sub 2}} (2.27+/-0.13l/h/l) occurred at C{sub s}=40g COD/l and HRT=2.2h, whereas the highest Y{sub H{sub 2}} (4.98+/-0.18mol H{sub 2}/mol sucrose) was obtained at C{sub s}=40g COD/l and HRT=8.9h. The correlation between the production rate and the organic loading rate (OLR) can be satisfactorily described by Monod-type models. There was no universal trend of the dependence between the H{sub 2} yield and OLR. The H{sub 2} content in the biogas was stably maintained at over 40%. The major soluble products were butyric acid and acetic acid, as they accounted for 62-73% and 16-22% of total soluble microbial products (SMPs), respectively. The H{sub 2}-producing performance in the DTFBR system can be stably maintained and reproducible in long-term operations, while unstable operations can be quickly recovered via proper thermal treatment at 70-80{sup o}C. (author)

  2. Influence of structure improvement of guide tubes and bundles in pressurized water reactor (PWR) on drop of control rods

    International Nuclear Information System (INIS)

    Shen Xiuzhong; Yu Pingan; Yang Guanyue

    1996-01-01

    In order to alleviate the cross hydraulic load on control rod guide tubes and bundles, some protective sleeves are added to those near the upper plenum outlet nozzles (4 symmetric bundles: 02-26, 03-25, 11-29, 12-28). In a 1/4 scale transparent model of the PWR upper plenum of Qinshan Nuclear Power Station, water was chosen as the fluid and hydraulic experiments with improved control rod guide tubes and bundles were carried out. The results were carefully compared with those of the experiments with unimproved control rod guide tubes and bundles. It is concluded that adding protective sleeves to the control rod guide tubes and bundles near the outlet nozzles will help to lighten the hydraulic load on them and make certain of the free movement and rapid dropping of control rods in the tubes and bundles in emergency by order

  3. Some approaches to numerical modelling of a phenomenon observed during steam generator tube rupture in the reactor with liquid metal coolant

    Directory of Open Access Journals (Sweden)

    Usov Eduard V.

    2017-01-01

    Full Text Available The presented paper contains a description of approaches to simulate processes, observed during leakage in the steam generator of the reactor with liquid metal coolant. These approaches have been implemented in thermal hydraulic code HYDRA-IBRAE/LM. To calculate motion of gas bubbles in liquid metal flow and heat transfer of gas bubbles with metal, different relations are used in HYDRA-IBRAE/LM code. The code contains models of chemical interaction between water and sodium for modelling of tube rupture in sodium cooled fast reactors, Modelling of the experiments has been made using HYDRA-IBRAE/LM code. The results of the modelling with determined main factors are presented in the article.

  4. The Virginia pine sawfly in 1960 - a special cooperative report

    Science.gov (United States)

    T. McIntyre; R. C. Heller; C. L. Morris

    1961-01-01

    An outbreak of the pine sawfly, Neodiprion pratti pratti (Dyar), has existed in Maryland since 1955. By 1959 the insect had spread throughout 14 million acres in the Coastal Plain and Piedmont of Virginia and into several North Carolina counties. Because egg surveys conducted in the spring of 1960 indicated a continuation of the epidemic, an aerial survey was conducted...

  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. 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)

    Macedo, Luiz Alberto; Torres, Walmir Maximo

    2009-01-01

    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)

  7. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1982-04-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1980. Tube defects occurred at 38% of the 97 reactors surveyed. This is a marginal improvement over 1979 when defects occurred at 41% of the reactors. The number of failed tubes was also lower, 0.14% of the tubes in service in 1980 compared with 0.20% of those in service in 1979. Analysis of the causes of these failures indicates that stress corrosion cracking was the leading failure mechanism. Reactors that used all-volatile treatment of secondary water, with or without full-flow condensate demineralization since start-up showed the lowest incidence of corrosion-related defects

  8. USE OF A GRIFFITH TUBE TO EVALUATE THE ANAEROBIC SLUDGE SEDIMENTATION IN A UASB REACTOR TREATING AN EFFLUENT WITH LONG-CHAIN FATTY ACIDS

    Directory of Open Access Journals (Sweden)

    L. A. S. Miranda

    Full Text Available Abstract This paper proposes to study the sedimentation characteristics of anaerobic sludge, by determining the settling velocity of sludge granules with the Griffith Tube. This is a simple, low-cost method, suitable for use in full-scale treatment plants. The settling characteristics of sludge from two laboratory-scale UASB reactors fed with saccharose and different concentrations of sodium oleate and sodium stereate were evaluated. Addition of fatty acids caused a gradual destabilization of the system, affecting overall performance. The sedimentation profile changed after addition of fatty acids to the synthetic substrate, decreased sedimentation velocity and increased granule diameter. This behaviour was attributed to the adsorption of fatty acids onto the granules, modifying the diameter, shape and density of these bioparticles.

  9. 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)

    Hummel, D.W.; Novog, D.R.

    2012-01-01

    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 PuO 2 in ThO 2 ) 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% PuO 2 (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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, D.W.; Novog, D.R. [McMaster Univ., Hamilton, Ontario (Canada)

    2012-07-01

    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 PuO{sub 2} in ThO{sub 2}) 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% PuO{sub 2} (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)

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

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1985-01-01

    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 - 0 F h) cooldown-rate limit was exceeded and AFW could not be injected. Also, the guidelines did not address the SG overfill issue

  12. Virginia: The Old Dominion

    Science.gov (United States)

    Kirschenmann, Jean

    2009-01-01

    In this article, the author features Virginia, a state of contrasts--from the suburbs of Washington, District of Columbia, in northern Virginia to the Great Dismal Swamp in the south and from the scenic mountaintops in the west to the beachside resorts along its eastern shore. Virginia became a state in 1788, the tenth state to join the Union.…

  13. 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.

  14. Steam generator tube failures

    International Nuclear Information System (INIS)

    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

  15. Effect of Loop Configuration on Steam Drum Level Control for a Multiple Drum Interconnected Loops Pressure Tube Type Boiling Water Reactor

    Science.gov (United States)

    Gaikwad, Avinash J.; Vijayan, P. K.; Iyer, Kannan; Bhartiya, Sharad; Kumar, Rajesh; Lele, H. G.; Ghosh, A. K.; Kushwaha, H. S.; Sinha, R. K.

    2009-12-01

    For AHWR (Advanced Heavy Water Reactor), a pressure tube type Boiling Water Reactor (BWR) with parallel inter-connected loops, the Steam Drum (SD) level control is closely related to Main Heat Transport (MHT) coolant inventory and sustained heat removal through natural circulation, hence overall safety of the power plant. The MHT configuration with multiple (four) interconnected loops influences the SD level control in a manner which has not been previously addressed. The MHT configuration has been chosen based on comprehensive overall design requirements and certain Postulated Initiated Event (PIEs) for Loss of Coolant Accident (LOCA), which postulates a double ended break in the four partitioned Emergency Core Cooling System (ECCS) header. A conventional individual three-element SD level controller can not account for the highly coupled and interacting behaviors, of the four SD levels. An innovative three-element SD level control scheme is proposed to overcome this situation. The response obtained for a variety of unsymmetrical disturbances shows that the SD levels do not diverge and quickly settle to the various new set points assigned. The proposed scheme also leads to enhanced safety margins for most of the PIEs considered with a little influence on the 100% full power steady-state design conditions.

  16. 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)

    Yoon, C.; Rhee, B. W.; Chung, B. D.; Cho, Y. J.; Kim, M. W.

    2008-01-01

    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

  17. Water-quality assessment of the Albermarle-Pamlico drainage basin, North Carolina and Virginia; a summary of selected trace element, nutrient, and pesticide data for bed sediments, 1969-90

    Science.gov (United States)

    Skrobialowski, S.C.

    1996-01-01

    Spatial distributions of metals and trace elements, nutrients, and pesticides and polychiorinated biphenyls (PCB's) in bed sediment were characterized using data collected from 1969 through 1990 and stored in the U.S. Geological Survey's National Water Data Storage and Retrieval (WATSTORE) system and the U.S. Environmental Protection Agency's Storage and Retrieval (STORET) system databases. Bed-sediment data from WATSTORE and STORET were combined to form a single database of 1,049 records representing 301 sites. Data were examined for concentrations of 16 metals and trace elements, 4 nutrients, 10 pesticides, and PCB's. Maximum bed-sediment concentrations were evaluated relative to sediment-quality guidelines developed by the National Oceanic and Atmospheric Administration, the Ontario Ministry of Environment and Energy, and the Virginia Department of Environmental Quality. Sites were not selected randomly; therefore, results should not be interpreted as representing average conditions. Many sites were located in or around lakes and reservoirs, urban areas, and areas where special investigations were conducted. Lakes and reservoirs function as effective sediment traps, and elevated concentrations of some constituents occurred at these sites. High concentrations of many metals and trace elements also occurred near urban areas where streams receive runoff or inputs from industrial, residential, and municipal activities. Elevated nutrient concentrations occurred near lakes, reservoirs, and the mouths of major rivers. The highest concentrations of arsenic, beryllium, chromium, iron. mercury, nickel, and selenium occurred in the Roanoke River Basin and may be a result of geologic formations or accumulations of bed sediment in lakes and reservoirs. The highest concentrations of cadmium, lead, and thallium were detected in the Chowan River Basin; copper and zinc were reported highest in the Neuse River Basin. Total phosphorus and total ammonia plus organic nitrogen

  18. New constraints on the late Cenozoic incision history of the New River, Virginia

    OpenAIRE

    Ward, Dylan J.

    2004-01-01

    The New River crosses the core of the ancient, tectonically quiescent Appalachian orogen as it follows its course through North Carolina, Virginia, and West Virginia. It is ideally situated to record the changes in geomorphic process rates that occur in the Appalachians as a response to late Cenozoic climate variations. Active erosion features on resistant bedrock that floors the river at prominent knickpoints demonstrate that the river is currently incising toward base level. However, large ...

  19. Virginia Forest Landowners Events Calendar

    OpenAIRE

    1998-01-01

    A quarterly calendar listing events that promote Virginia forest stewardship through sustainable forestry, wildlife management, timber marketing, outdoor recreation, and soil and water conservation. Calendar sponsors include: Virginia Forestry Association (VFA); VFA Sustainable Forestry Task Force Virginia Department of Forestry; Virginia Forestry Educational Foundation; VA Tech College of Forestry & Wildlife Resources; Virginia Cooperative Extension

  20. Virginia Forest Landowners Events Calendar

    OpenAIRE

    1997-01-01

    A quarterly calendar listing events that promote Virginia forest stewardship through sustainable forestry, wildlife management, timber marketing, outdoor recreation, and soil and water conservation. Calendar sponsors include: Virginia Forestry Association (VFA); VFA Sustainable Forestry Task Force Virginia Department of Forestry; Virginia Forestry Educational Foundation; VA Tech College of Forestry & Wildlife Resources; Virginia Cooperative Extension

  1. PENGARUH SUHU DAN LAMA PROSES SULFONASI DALAM PROSES PRODUKSI METHYL ESTER SULFONIC ACID (MESA MENGGUNAKAN SINGLE TUBE FALLING FILM REACTOR (STFR

    Directory of Open Access Journals (Sweden)

    Siti Mujdalipah

    2013-03-01

    Full Text Available Effects of Temperature and Sulfonation Time on Methyl Ester Sulfonic Acid (MESA Production Process usingSingle Tube Falling Film Reactor (STFR Siti Mujdalipah, Erliza Hambali, Ani Suryani, Edi Zulchaidir ABSTRAK Methyl Ester Sulfonic Acid (MESA merupakan produk antara dari surfaktan Metil Ester Sulfonat (MES. MESmemiliki beragam aplikasi dalam produk personal care, pencuci dan pembersih, dan untuk Enhanced Oil Recovery(EOR. Proses produksi MESA menggunakan gas SO3 dalam Single Tube Falling Film Reactor (STFR merupakanteknologi yang umum digunakan. Kajian ini bertujuan untuk mendapatkan kondisi proses sulfonasi metil ester oleinterbaik menggunakan gas SO3 dalam STFR. Kajian dilakukan dalam tiga tahap, yaitu tahap penelitian, tahap analisis,dan tahap pengolahan data. Tahap produksi MESA terdiri dari pembuatan metil ester (ME dari olein minyak sawit dankajian pengaruh suhu dan lama proses sulfonasi. Tahap analisis meliputi analisis sifat Þ siko kimia olein minyak sawit,analisa sifat Þ siko kimia ME olein sawit, dan analisis sifat Þ siko kimia MESA olein sawit. Kajian pengaruh suhu danlama proses sulfonasi terhadap proses sulfonasi metil ester olein terdiri dari suhu 70, 90, dan 110 oC dan lama prosessulfonasi 30, 60, dan 90 menit. Analisis varian pada !=0,01 menunjukan bahwa lama proses sulfonasi berpengaruh nyataterhadap kadar bahan aktif. Analisis varian pada !=0,01 juga menunjukan bahwa lama proses sulfonasi berpengaruhnyata terhadap nilai pH, bilangan asam, bilangan iod, dan kemampuan MESA dalam menurunkan tegangan antarmuka(IFT, Interfacial Tension antara air formasi dan minyak bumi. Proses sulfonasi terbaik dicapai pada suhu sulfonasi 90oCdan lama proses sulfonasi 90 menit. Kondisi proses sulfonasi terbaik dapat menghasilkan MESA dengan karakteristikkadar bahan aktif 31,44%, pH 2,66, bilangan asam 24,88 ml NaOH/g sampel, bilangan iod 11,95 mg I/g sampel, danmemiliki kemampuan menurunkan IFT antara air formasi dan minyak bumi dari 30 dyne

  2. Effects of heat exchanger tubes on hydrodynamics and CO 2 capture of a sorbent-based fluidized bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lai, Canhai; Xu, Zhijie; Li, Tingwen; Lee, Andrew; Dietiker, Jean-François; Lane, William; Sun, Xin

    2017-12-01

    In virtual design and scale up of pilot-scale carbon capture systems, the coupled reactive multiphase flow problem must be solved to predict the adsorber’s performance and capture efficiency under various operation conditions. This paper focuses on the detailed computational fluid dynamics (CFD) modeling of a pilot-scale fluidized bed adsorber equipped with vertical cooling tubes. Multiphase Flow with Interphase eXchanges (MFiX), an open-source multiphase flow CFD solver, is used for the simulations with custom code to simulate the chemical reactions and filtered models to capture the effect of the unresolved details in the coarser mesh for simulations with reasonable simulations and manageable computational effort. Previously developed two filtered models for horizontal cylinder drag, heat transfer, and reaction kinetics have been modified to derive the 2D filtered models representing vertical cylinders in the coarse-grid CFD simulations. The effects of the heat exchanger configurations (i.e., horizontal or vertical) on the adsorber’s hydrodynamics and CO2 capture performance are then examined. The simulation result subsequently is compared and contrasted with another predicted by a one-dimensional three-region process model.

  3. Prediction of diametral creep for pressure tubes of a pressurized heavy water reactor using data based modeling

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Yong [Central Research Institute, Korea Hydro and Nuclear Power Company, Daejeon (Korea, Republic of); Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2012-05-15

    The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict Pressure Tube (PT) diametral creep employing the previously measured PT diameters and operating conditions. There are twelve bundles in a fuel channel, and for each bundle a linear model was developed by using the dependent variables, such as the fast neutron fluences and the bundle coolant temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3, and 4 of the Wolsung nuclear plant in Korea were used to develop the BPLM. The data from the remaining 10 channels were used to test the developed BPLM. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from Units 2, 3, and 4. Two error components for the BPLM, which are the epistemic error and the aleatory error, were generated. The diametral creep prediction and two error components will be used for the generation of the regional overpower trip setpoint at the corresponding effective full power days. The root mean square (RMS) errors were also generated and compared to those from the current prediction method. The RMS errors were found to be less than the previous errors.

  4. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  5. Forests of Virginia, 2016

    Science.gov (United States)

    T.J. Brandeis; A.J. Hartsell; K.C. Randolph; C.M. Oswalt

    2018-01-01

    This resource update provides an overview of forest resources in Virginia based on an inventory conducted by the U.S. Forest Service, Forest Inventory and Analysis (FIA) program at the Southern Research Station in cooperation with the Virginia Department of Forestry.

  6. Forests of Virginia,2013

    Science.gov (United States)

    Anita K. Rose

    2015-01-01

    This resource update provides an overview of forest resources in Virginia. Information for this factsheet was updated by means of the Forest Inventory and Analysis (FIA) annualized sample design. Each year, 20 percent of the sample plots (one panel) in Virginia are measured by field crews, the data compiled, and new estimates produced.

  7. REACTOR SHIELD

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  8. Steam generator tube vibration study

    International Nuclear Information System (INIS)

    Enderlin, W.I.

    1986-01-01

    Chemical cleaning has been proposed to remove magnetite buildup in some pressurized water reactor steam generators. The US Nuclear Regulatory Commission (NRC) has expressed concern that such cleaning would combine with the tube denting caused by magnetite formation to enlarge tube/tube support plate clearances, increasing the level of flow-induced vibrations that could lead to unacceptably high tube wear and failure rates. In support of NRC, the Pacific Northwest Laboratory investigated whether such increased clearances would exacerbate tube fretting wear. Using a full-length scale model of a steam generator tube bundle, flow tests were conducted at an instrumented location through clearances representing as-built and post-cleaned tube conditions. Test results indicated little potential for increased tube wear as a result of chemical cleaning, under normal operating conditions at tube support locations similar to that tested

  9. SLOWPOKE reactor

    International Nuclear Information System (INIS)

    Evans, D.J.R.; Downs, W.E.

    1974-01-01

    The SLOWPOKE reactor is described, which is a small pool type with thermal neutron fluxes ranging from 10 11 -10 12 n cm -2 sec -1 . It differs in many ways from conventional pool type, namely small critical mass, beryllium reflector and a closed reactor container. The reactor is designed as small and simply as possible, and consistently with safety and good operating practice. Access to the present model is via pneumatic irradiation tubes only, which limits the use of the facility to activation analysis, tracer production and training. (Mori, K.)

  10. Litter survey in Virginia.

    Science.gov (United States)

    1976-01-01

    This report summarizes the findings of the litter survey for highways, urban areas, and recreational areas as specified in the "Virginia Litter Control Act". Litter samples from 61 highway sites, 11 urban sites, and 10 recreational sites geographical...

  11. 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.

  12. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    1975-01-01

    A description is given of a cooling water intake collector for a nuclear reactor. It includes multiple sub-collectors extending out in a generally parallel manner to each other, each one having a first end and a second one separated along their length, and multiple water outlets for connecting each one to a corresponding pressure tube of the reactor. A first end tube and a second one connect the sub-collector tubes together to their first and second ends respectively. It also includes multiple collector tubes extending transversely by crossing over the sub-collector tubes and separated from each other in the direction of these tubes. Each collector tubes has a water intake for connecting to a water pump and multiple connecting tubes separated over its length and connecting each one to the corresponding sub-collector [fr

  13. Prospects for stronger calandria tubes

    International Nuclear Information System (INIS)

    Ells, C.E.; Coleman, C.E.; Hosbons, R.R.; Ibrahim, E.F.; Doubt, G.L.

    1990-12-01

    The CANDU calandria tubes, made of seam welded and annealed Zircaloy-2, have given exemplary service in-reactor. Although not designed as a system pressure containment, calandria tubes may remain intact even in the face of pressure tube rupture. One such incident at Pickering Unit 2 demonstrated the economic advantage of such an outcome, and a case can be made for increasing the probability that other calandria tubes would perform in a similar fashion. Various methods of obtaining stronger calandria tubes are available, and reviewed here. When the tubes are internally pressurized, the weld is the weak section of the tube. Increasing the oxygen concentration in the starting sheet, and thickening the weld, are promising routes to a stronger tube

  14. Earthquakes in South Carolina and Vicinity 1698-2009

    Science.gov (United States)

    Dart, Richard L.; Talwani, Pradeep; Stevenson, Donald

    2010-01-01

    This map summarizes more than 300 years of South Carolina earthquake history. It is one in a series of three similar State earthquake history maps. The current map and the previous two for Virginia and Ohio are accessible at http://pubs.usgs.gov/of/2006/1017/ and http://pubs.usgs.gov/of/2008/1221/. All three State earthquake maps were collaborative efforts between the U.S. Geological Survey and respective State agencies. Work on the South Carolina map was done in collaboration with the Department of Geological Sciences, University of South Carolina. As with the two previous maps, the history of South Carolina earthquakes was derived from letters, journals, diaries, newspaper accounts, academic journal articles, and, beginning in the early 20th century, instrumental recordings (seismograms). All historical (preinstrumental) earthquakes that were large enough to be felt have been located based on felt reports. Some of these events caused damage to buildings and their contents. The more recent widespread use of seismographs has allowed many smaller earthquakes, previously undetected, to be recorded and accurately located. The seismicity map shows historically located and instrumentally recorded earthquakes in and near South Carolina

  15. Lung retention and clearance classification of a 14C-containing aerosol produced during re-tubing of a nuclear reactor

    International Nuclear Information System (INIS)

    Johnson, J.R.

    1989-01-01

    Following the discovery of a 14 C-bearing aerosol during the re-tubing of a CANDU reactor, a procedure was developed to monitor workers for inhaled 14 C by measuring the amount of 14 C excreted in feces. This procedure requires that an appropriate lung model be used to give the relationship between the fecal excretion and 14 C in lung and possibly other tissues. Since preliminary studies on samples from air filters indicated that that the 14 C aerosol was insoluble, it was decided to use the International Commission on Radiological Protection's (ICRP) lung model for insoluble material, commonly called Class Y (ICRP 1966; ICRP 1979). In order to ensure that the selection of the ICRP Class Y lung model was appropriate for this aerosol, an experiment was carried out to measure the rate that 14 C was cleared from the deep lung to the GI tract and then excreted in feces, and the rate that 14 C was solubilized and transferred to liver or excreted in urine in the rat. Liver and urinary excretion were chosen because it was assumed that any 14 C transferred to blood would be in an unknown organic form, and hence taken up the liver, or would be as HCO 3 - /CO 2 . It is expected that both of these possibilities would result in 14 C in urea in the urine (Orten and Neuhaus 1986). Extrapolations to humans can be made by comparison of the 14 C results to those for a 141 CeO 2 control exposure carried out at the same time. The CeO 2 was selected as the control exposure material as it was readily available and had been used by other laboratories (e.g., Lundgren et al. 1980). Its retention and excretion is consistent with that described by the ICRP Class Y model. Hence, if the 14 C aerosol had a retention and excretion pattern similar to CeO 2 , its retention and excretion would also be reasonably described by the Class Y model. This note summarizes the results of this experiment. (author)

  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. Cogeneration and North Carolina

    Energy Technology Data Exchange (ETDEWEB)

    Kohl, J. [ed.

    1979-01-01

    A separate abstract was prepared for each of 18 individual presentations. Appendices include lists of participants, speakers, and session chairmen plus California and North Carolina reports and legislation dealing with cogeneration.

  18. Feeding Tubes

    Science.gov (United States)

    ... the TPN. Tubes Used for Enteral Feeds NG (Nasogastric Tube) A flexible tube is placed via the nose, ... portion of the small intestine Naso – nose NG – Nasogastric Tube -ostomy – new opening Percutaneous – through the skin PEJ – ...

  19. Ceramic oxygen transport membrane array reactor and reforming method

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.; Christie, Gervase Maxwell; Robinson, Charles; Wilson, Jamie R.; Gonzalez, Javier E.; Doraswami, Uttam R.

    2016-11-08

    The invention relates to a commercially viable modular ceramic oxygen transport membrane reforming reactor configured using repeating assemblies of oxygen transport membrane tubes and catalytic reforming reactors.

  20. Liquid metal fast breeder reactor steam generator: behaviour of heat exchange tubes in face of a through crack resulting in a contact between sodium and water

    International Nuclear Information System (INIS)

    Quinet, J.L.; Lannou, L.

    1978-01-01

    The results of a survey made Electricite de France on the behaviour of cracked tubes under operating conditions of an industrial steam generator are submitted in this communication. A comparison is made of the tube material: INCOLOY 800, 2 1/4 Cr-1 Mo, 9 Cr-2 Mo land to the initial leak. Finally, a description is given of the self-development process of a water leak into sodium. (author)

  1. Library Programs in North Carolina

    Data.gov (United States)

    Town of Chapel Hill, North Carolina — Count of programs offered and program attendance numbers at public libraries in North CarolinaData is from the 2014-15 NC Statistical Report of NC Public Libraries:...

  2. Quaternary geophysical framework of the northeastern North Carolina coastal system

    Science.gov (United States)

    Thieler, E.R.; Foster, D.S.; Mallinson, D.M.; Himmelstoss, E.A.; McNinch, J.E.; List, J.H.; Hammar-Klose, E.S.

    2013-01-01

    The northeastern North Carolina coastal system, from False Cape, Virginia, to Cape Lookout, North Carolina, has been studied by a cooperative research program that mapped the Quaternary geologic framework of the estuaries, barrier islands, and inner continental shelf. This information provides a basis to understand the linkage between geologic framework, physical processes, and coastal evolution at time scales from storm events to millennia. The study area attracts significant tourism to its parks and beaches, contains a number of coastal communities, and supports a local fishing industry, all of which are impacted by coastal change. Knowledge derived from this research program can be used to mitigate hazards and facilitate effective management of this dynamic coastal system.

  3. Nuclear reactor with control rods

    International Nuclear Information System (INIS)

    Obermeyer, F.D.; Berringer, R.T.

    1979-01-01

    A liquid-cooled nuclear reactor including fuel assemblies mounted within a reactor vessel having linearly movable control rods passing through control rod guide tubes into respective aligned fuel assemblies is described. Reactor coolant circulates through the assemblies. Guide tubes and other vessel internals structures located above the assemblies and is discharged through an outlet nozzle positioned above the elevation of primary flow openings in the guide tube walls. The guide tube includes internal horizontal supports and a length limited continuous control rod guide which, in conjunction with the flow openings, alleviate detrimental coolant cross flows and frictional restraints imposed upon the control rods

  4. DEVELOPMENT OF AN ANALYTICAL METHOD TO EVALUATE THE INTEGRITY OF A CALANDRIA TUBE IN THE CASE OF PRESSURE TUBE RUPTURE

    OpenAIRE

    森下 善嗣

    1990-01-01

    An analytical method which consists of two‐dimensional thermal‐hydraulic analysis and three‐dimensional structural analysis has been proposed to evaluate the integrity of a calandria tube in the case of pressure tube rupture in a pressure tube type reactor.In order to validate the method,experiments were carried out with coaxially arranged double tubes simulating a pressure tube and a calandria tube.Experimental data were also comparedwith analytical results with the proposed method.

  5. Measurement device for reactor

    International Nuclear Information System (INIS)

    Sakamoto, M.

    1982-02-01

    A measurement device for a reactor is described. It consists of closed end guide tubes positioned vertically beneath each of the fuel assemblies; the ends of these tubes are immersed in the core coolant fluid. A ''free space'' in-pile detector and a detection device are enclosed in each of the guide tubes. The state of the core is characterized by output signals delivered by the in-pile detectors. These detectors are of the acoustic type [fr

  6. 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.

  7. Libraries in Virginia: MedlinePlus

    Science.gov (United States)

    ... https://medlineplus.gov/libraries/virginia.html Libraries in Virginia To use the sharing features on this page, please enable JavaScript. Charlottesville University of Virginia Health System Claude Moore Health Sciences Library 1350 ...

  8. In situ sampling for pressure tube deuterium concentration

    International Nuclear Information System (INIS)

    Harrington, A.J.; Kittmer, C.A.

    1988-01-01

    The present method of assessing the useful life of pressure tubes in CANDU (CANada Deuterium Uranium) reactors requires the periodic removal and examination of a tube. Special tooling was developed at Atomic Energy of Canada Limited (AECL) to obtain a sample of material from a pressure tube without removing the tube from the reactor. The sampling tool concept has been successfully used by Ontario Hydro during scheduled outages at the Pickering Nuclear Generating Station (PNGS). (author)

  9. West Virginia Forests 2013

    Science.gov (United States)

    Randall S. Morin; Gregory W. Cook; Charles J. Barnett; Brett J. Butler; Susan J. Crocker; Mark A. Hatfield; Cassandra M. Kurtz; Tonya W. Lister; William G. Luppold; William H. McWilliams; Patrick D. Miles; Mark D. Nelson; Charles H. (Hobie) Perry; Ronald J. Piva; James E. Smith; Jim Westfall; Richard H. Widmann; Christopher W. Woodall

    2016-01-01

    The annual inventory of West Virginia's forests, completed in 2013, covers nearly 12.2 million acres of forest land with an average volume of more than 2,300 cubic feet per acre. This report is based data collected from 2,808 plots located across the State. Forest land is dominated by the oak/hickory forest-type group, which occupies 74 percent of total forest...

  10. West Virginia's Forests 2008

    Science.gov (United States)

    Richard H. Widmann; Gregory W. Cook; Charles J. Barnett; Brett J. Butler; Douglas M. Griffith; Mark A. Hatfield; Cassandra M. Kurtz; Randall S. Morin; W. Keith Moser; Charles H. Perry; Ronald J. Piva; Rachel Riemann; Christopher W. Woodall

    2012-01-01

    The first full annual inventory of West Virginia's forests reports 12.0 million acres of forest land or 78 percent of the State's land area. The area of forest land has changed little since 2000. Of this land, 7.2 million acres (60 percent) are held by family forest owners. The current growing-stock inventory is 25 billion cubic feet--12 percent more than in...

  11. Carolinas Communication Annual, 1998.

    Science.gov (United States)

    McLennan, David B.

    1998-01-01

    This 1998 issue of "Carolinas Communication Annual" contains the following articles: "Give Me That Old Time Religion?: A Study of Religious Themes in the Rhetoric of the Ku Klux Klan" (John S. Seiter); "The Three Stooges versus the Third Reich" (Roy Schwartzman); "Interdisciplinary Team Teaching: Implementing…

  12. 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.

  13. Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

    International Nuclear Information System (INIS)

    Bucholz, J.A.

    2000-01-01

    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

  14. Power control device for nuclear reactors

    International Nuclear Information System (INIS)

    Kagawa, Tatsuo

    1984-01-01

    Purpose: To eliminate for requirement of control rods and movable portions, as well as ensure the safety and reliability of the operation. Constitution: A plurality of control tubes are disposed within a reactor core instead of control rods. Tubes are connected from below the reactor core to the control tubes for supplying liquid poisons such as aqueous boric acid to the inside of the control tubes. Further, tubes are connected to the upper portion of the control tubes for guiding the liquid poisons from the reactor core to the outside. The tubes for supplying and discharging the liquid poisons are introduced externally through the flange disposed at the upper portion of a pressure vessel. At the outside of the pressure vessel, are disposed a liquid poison tank, a pressurizing source, a pressure control valve, a liquid level meter and the like. The control for the reactor power is conducted by controlling the level of the liquid poisons in the control tubes. (Ikeda, J.)

  15. Reactor shutdown method

    International Nuclear Information System (INIS)

    Nishino, Yoshitaka; Sawa, Toshio; Matsumoto, Takayuki; Osumi, Katsumi; Usui, Naoshi.

    1991-01-01

    A device for injecting a hydrogen gas, a chelating agent or a reducing agent is disposed in a reactor water recycling system. Upon reactor shutdown, the hydrogen gas, the chelating agent or the reducing agent is injected to primary coolants. With such a procedure, radioactive ions formed by the dissolution of oxide layers at the surface of pipelines and equipments in a reactor water recycling system and a reactor water cleanup system are removed from the primary coolants by a reactor water cleanup device. Accordingly, since the dose rate at the surface of the pipelines can be reduced, the operator's radiation dose can be reduced upon periodical inspection for a power plant. Further, the inner pressure of the reactor is kept higher than the saturated steam pressure at the reactor water temperature to suppress boiling of the reactor water. This can suppress the peeling of cruds deposited to the surface of the fuel cladding tube. (I.N.)

  16. Nuclear reactor containment device

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu.

    1980-01-01

    Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

  17. The isolated red spruce communities of Virginia and West Virginia

    Science.gov (United States)

    Harold S. Adams; Steven Stephenson; Adam W. Rollins; Mary Beth. Adams

    2010-01-01

    Quantitative data on the composition and structure of coniferous forests containing red spruce in the mountains of central and southwestern Virginia and eastern central West Virginia, based on sampling carried out in 67 stands during the 1982 to 1984 field seasons, are provided. The average importance value ([relative basal area + relative density/2]) of red spruce was...

  18. Integrated nuclear reactor

    International Nuclear Information System (INIS)

    Pales, I.; Hasko, V.

    1984-01-01

    The reactor is provided with an integrated circuit of primary medium circulation with hydraulic pump drive. The pump drive which is a blade hydraulic facility is placed in the reactor vessel together with the pump. The primary medium flows through the core and enters the inter-tube space of the secondary circuit heat exchanger. The secondary circuit medium is supplied under the bottom tube plate with a supply pipe. From it the flow of secondary medium is directed to the blades of the hydraulic facility, e.g. the turbine. The turbine drives the pump which transports the primary medium to the reactor core. The secondary medium enters the heat exchanger tubes and through their walls receives the heat from the primary medium. This design reduces capital costs of the reactor and increases its safety. (E.S.)

  19. Nuclear reactor spacer assembly

    International Nuclear Information System (INIS)

    Anthony, A.J.; Groves, M.D.

    1979-01-01

    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

  20. 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.

  1. Fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Gjertsen, R.K.; Tower, S.N.; Huckestein, E.A.

    1982-01-01

    A fuel assembly for a nuclear reactor comprises a 5x5 array of guide tubes in a generally 20x20 array of fuel elements, the guide tubes being arranged to accommodate either control rods or water displacer rods. The fuel assembly has top and bottom Inconel (Registered Trade Mark) grids and intermediate Zircaloy grids in engagement with the guide tubes and supporting the fuel elements and guide tubes while allowing flow of reactor coolant through the assembly. (author)

  2. Analysis of the Reduction Rate of Hematite Concentrate Particles in the Solid State by H2 or CO in a Drop-Tube Reactor Through CFD Modeling

    Science.gov (United States)

    Fan, De-Qiu; Sohn, H. Y.; Elzohiery, Mohamed

    2017-10-01

    The kinetic analysis of the reduction of hematite concentrate particles by individual reducing gas H2 or CO was performed using a computational fluid dynamics (CFD)-based approach in this paper. The particle residence time was calculated through the integration of the equation of particle motion. Non-uniform particle temperature profiles inside the reactor were obtained, and were taken into consideration for the kinetic analysis. The calculated reduction degrees based on this approach are in good agreement with the experimental values.

  3. Reactor and method for production of nanostructures

    Science.gov (United States)

    Sunkara, Mahendra Kumar; Kim, Jeong H.; Kumar, Vivekanand

    2017-04-25

    A reactor and method for production of nanostructures, including metal oxide nanowires or nanoparticles, are provided. The reactor includes a regulated metal powder delivery system in communication with a dielectric tube; a plasma-forming gas inlet, whereby a plasma-forming gas is delivered substantially longitudinally into the dielectric tube; a sheath gas inlet, whereby a sheath gas is delivered into the dielectric tube; and a microwave energy generator coupled to the dielectric tube, whereby microwave energy is delivered into a plasma-forming gas. The method for producing nanostructures includes providing a reactor to form nanostructures and collecting the formed nanostructures, optionally from a filter located downstream of the dielectric tube.

  4. Forests of West Virginia, 2013

    Science.gov (United States)

    Richard H. Widmann

    2014-01-01

    This publication provides an overview of the forest resources in West Virginia based upon inventories conducted by the U.S. Forest Service, Forest Inventory and Analysis (FIA) program of the Northern Research Station. Information about the FIA program is available online at http://fia.fs.fed.us. Since 2004, FIA has implemented an annual inventory in West Virginia. For...

  5. SODIUM DEUTERIUM REACTOR

    Science.gov (United States)

    Oppenheimer, E.D.; Weisberg, R.A.

    1963-02-26

    This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

  6. Hydrogeology of Virginia

    Science.gov (United States)

    Nelms, David L.; Harlow, George; Bruce, T. Scott; Bailey, Christopher M.; Sherwood, W. Cullen; Eaton, L. Scott; Powars, David S.

    2016-01-01

    The hydrogeology of Virginia documented herein is in two parts. Part 1 consists of an overview and description of the hydrogeology within each regional aquifer system in the Commonwealth. Part 2 includes discussions of hydrogeologic research topics of current relevance including: 1. the Chesapeake Bay impact structure, 2. subsidence/compaction in the Coastal Plain, 3. groundwater age and aquifer susceptibility, 4. the occurrence of groundwater at depth in fractured-rock and karst terrains, and 5. hydrologic response of wells to earthquakes around the world.

  7. The 2001 Virginia Rural Homeless Survey

    OpenAIRE

    Koebel, C. Theodore; Murphy, Michelle; Brown, Adam

    2001-01-01

    The Virginia Center for Housing Research was commissioned by the Virginia Housing Study Commission, the Virginia Interagency Action Council for the Homeless, and the Virginia Department of Housing and Community Development to conduct this research in response to House Joint Resolution 257 requesting a study of the number and needs of homeless people living in rural areas of the Commonwealth.

  8. 21 CFR 808.98 - West Virginia.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false West Virginia. 808.98 Section 808.98 Food and... and Local Exemptions § 808.98 West Virginia. (a) The following West Virginia medical device... has exempted them from preemption: West Virginia Code, sections 30-26-14 (b) and (c) and section 30-26...

  9. Tube in shell heat exchangers

    International Nuclear Information System (INIS)

    Hayden, O.; Willby, C.R.; Sheward, G.E.; Ormrod, D.T.; Firth, G.F.

    1980-01-01

    An improved tube-in-shell heat exchanger to be used between liquid metal and water is described for use in the liquid metal coolant system of fast breeder reactors. It is stated that this design is less prone to failures which could result in sodium water reactions than previous exchangers. (UK)

  10. Assessment of the Polyacrylic Acid for an Ammonia Water Treatment and for Alloy 800NG SG Tube Material in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Lamouroux, Christine; You, Dominique; Plancque, Gabriel; Roy, Marc; Laire, Charles; Schnongs, Philippe

    2012-09-01

    To prevent the Steam Generators (SG) fouling by corrosion products or the Tube Support Plate (TSP) blockage the on-line injection of a dispersant such the Polyacrylic Acid (PAA) could be a relevant water treatment. Long-term trials performed in PWRs have shown that the PAA, injected at the SG inlet, facilitate the evacuation of the iron oxides by the SG blowdown. Given the ammonia treatment of the secondary water of the Belgian PWRs, the R and D program carried out was devoted to: - Verify the innocuousness of the PAA and its degradation products versus Alloy 800NG SCC susceptibility in case of over concentrations and sludge presence, - Assess the potential impact of the PAA and its thermal degradation products on the specific NH 3 water treatment. The main results can be summarized as following: The corrosion tests performed with PAA in case of over concentrations and sludge couldn't point out any negative effect of the dispersant on the SCC susceptibility of tubing materials such as Alloy 800NG. No significant modification of the tube oxide layer has been observed. At the SG operating temperature, the PAA is decomposed and a large spectrum from high to lower molecular weights polymers than the initial PAA arises. The fragmentation of the polymer into low molecular weight polyacrylic acids is obtained within 20 minutes and the average molecular weight is reduced by 50% from the original one. The thermal degradation products, their quantity and their kinetic of appearance, have been determined. The generated acetate concentration during the on-line dispersant application should remain low compared to the current values observed in the SG water. From the numerical simulation based on acetate concentration and on the kinetic law deduced from the experimental work, it can be concluded that in a 2-phase medium, the margin on the water pH compared to the neutral pH remains high. At 180 deg. C, no impact on the water pH is identified, taking into account realistic

  11. On the distribution of temperatures in steam generator tubes at tube support plate Intersections

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.; Petelin, S.

    1995-01-01

    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)

  12. Creep-fatigue design studies for a sodium-cooled fast reactor with tube sheet-to shell structure subjected to elevated temperature service

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, Jae Han [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-03-15

    In this paper, creep-fatigue damage under elevated temperatures is investigated for a tube sheet-to-shell structure, which is one of the main structures under Gen-IV class 1 components. To do this, detailed step-by-step procedures, including the elastic structural analysis and the ASME-NH code application, are described for a defined representative load cycle. From the sensitivity studies for various design parameters, such as hold time duration, shell thickness, and operating temperature, it is found that a reduction of thickness can decrease the thermal bending stresses, but the negative effect is that it may increase the primary stress and enhance the creep damage. The normal operating temperature is the most significant parameter in the creep-fatigue design

  13. Advanced pressure tube sampling tools

    International Nuclear Information System (INIS)

    Wittich, K.C.; King, J.M.

    2002-01-01

    Deuterium concentration is an important parameter that must be assessed to evaluate the Fitness for service of CANDU pressure tubes. In-reactor pressure tube sampling allows accurate deuterium concentration assessment to be made without the expenses associated with fuel channel removal. This technology, which AECL has developed over the past fifteen years, has become the standard method for deuterium concentration assessment. AECL is developing a multi-head tool that would reduce in-reactor handling overhead by allowing one tool to sequentially sample at all four axial pressure tube locations before removal from the reactor. Four sets of independent cutting heads, like those on the existing sampling tools, facilitate this incorporating proven technology demonstrated in over 1400 in-reactor samples taken to date. The multi-head tool is delivered by AECL's Advanced Delivery Machine or other similar delivery machines. Further, AECL has developed an automated sample handling system that receives and processes the tool once out of the reactor. This system retrieves samples from the tool, dries, weighs and places them in labelled vials which are then directed into shielded shipping flasks. The multi-head wet sampling tool and the automated sample handling system are based on proven technology and offer continued savings and dose reduction to utilities in a competitive electricity market. (author)

  14. Considerations in selecting tubing materials for CANDU steam generators

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    1978-01-01

    Corrosion resistance is the major consideration in selecting tubing material for CANDU steam generators. Corrosion, and additional considerations, lead to the following steam generator tubing material recommendations: for CANDU-BPHWR's (boiling pressurized heavy water reactors) low-cobalt Incoloy-800; for CANDU-PHWR's (pressurized, non-boiling, heavy water reactors), low-cobalt Monel-400

  15. Forests of North Carolina, 2013

    Science.gov (United States)

    Mark J. Brown

    2015-01-01

    This periodic resource update provides an overview of forest resources in North Carolina based on an inventory conducted by the U.S. Forest Service, Forest Inventory and Analysis (FIA) program at the Southern Research Station in cooperation with the North Carolina Forest Service. Data estimates are based on field data collected using the FIA annualized sample design...

  16. Forests of North Carolina, 2014

    Science.gov (United States)

    Mark Brown; Samuel Lambert

    2016-01-01

    This periodic resource update provides an overview of forest resources in North Carolina based on an inventory conducted by the U.S. Forest Service, Forest Inventory and Analysis (FIA) program at the Southern Research Station in cooperation with the North Carolina Forest Service. Data estimates are based on field data collected using the FIA annualized sample design...

  17. Virginia ESI: HABITATS (Habitat Polygons)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains sensitive biological resource data for submerged aquatic vegetation (SAV) and rare terrestrial plants/communities in Virginia. Vector polygons...

  18. Geothermal investigations in West Virginia

    Energy Technology Data Exchange (ETDEWEB)

    Hendry, R.; Hilfiker, K.; Hodge, D.; Morgan, P.; Swanberg, C.; Shannon, S.S. Jr.

    1982-11-01

    Deep sedimentary basins and warm-spring systems in West Virginia are potential geothermal resources. A temperature gradient map based on 800 bottom-hole temperatures for West Virginia shows that variations of temperature gradient trend northeasterly, parallel to regional structure. Highest temperature gradient values of about 28/sup 0/C/km occur in east-central West Virginia, and the lowest gradients (18/sup 0/C/km) are found over the Rome Trough. Results from ground-water geochemistry indicate that the warm waters circulate in very shallow aquifers and are subject to seasonal temperature fluctuations. Silica heat-flow data in West Virginia vary from about 0.89 to 1.4 HFU and generally increase towards the west. Bouguer, magnetic, and temperature gradient profiles suggest that an ancient rift transects the state and is the site of several deep sedimentary basins.

  19. 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...

  20. West Virginia 511 feasibility study.

    Science.gov (United States)

    2011-06-01

    Procedure for requesting a copy of the full report : Please submit your request, in writing, directly to the contact provided below. : Director of the Traffic Engineering Division : West Virginia Department of Transportation, Division of Highways : B...

  1. 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...

  2. Virginia Bridge Information Systems Laboratory.

    Science.gov (United States)

    2014-06-01

    This report presents the results of applied data mining of legacy bridge databases, focusing on the Pontis and : National Bridge Inventory databases maintained by the Virginia Department of Transportation (VDOT). Data : analysis was performed using a...

  3. 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 data...

  4. Assessment and management of ageing of major nuclear power plant components important to safety: CANDU pressure tubes

    International Nuclear Information System (INIS)

    1998-08-01

    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

  5. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  6. Simulated nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Berta, V.T.

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end

  7. 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.

  8. Reactor container

    International Nuclear Information System (INIS)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-01-01

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.)

  9. Reactor container

    Energy Technology Data Exchange (ETDEWEB)

    Oikawa, Hirohide; Otonari, Jun-ichiro; Tozaki, Yuka.

    1993-09-07

    Partition walls are disposed between a reactor pressure vessel and a suppression chamber to separate a dry well to an upper portion and a lower portion. A communication pipe is disposed to the partition walls. One end of the communication pipe is opened in an upper portion of the dry well at a position higher than a hole disposed to a bent tube of the suppression chamber. When coolants overflow from a depressurization valve by an erroneous operation of an emergency reactor core cooling device, the coolants accumulate in the upper portion of the dry well. When the pipeline is ruptured at the upper portion of the pressure vessel, only the inside of the pressure vessel and the upper portion of the dry well are submerged in water. In this case, the water level of the coolants does not elevate to the opening of the commuication pipe but they flow into the suppression chamber from the hole disposed to the bent tube. Since the coolants do not flow out to the lower portion of the dry well, important equipments such as control rod drives disposed at the lower portion of the dry wall can be prevented from submerging in water. (I.N.).

  10. Spring/dimple instrument tube restraint

    Science.gov (United States)

    DeMario, E.E.; Lawson, C.N.

    1993-11-23

    A nuclear fuel assembly for a pressurized water nuclear reactor has a spring and dimple structure formed in a non-radioactive insert tube placed in the top of a sensor receiving instrumentation tube thimble disposed in the fuel assembly and attached at a top nozzle, a bottom nozzle, and intermediate grids. The instrumentation tube thimble is open at the top, where the sensor or its connection extends through the cooling water for coupling to a sensor signal processor. The spring and dimple insert tube is mounted within the instrumentation tube thimble and extends downwardly adjacent the top. The springs and dimples restrain the sensor and its connections against lateral displacement causing impact with the instrumentation tube thimble due to the strong axial flow of cooling water. The instrumentation tube has a stainless steel outer sleeve and a zirconium alloy inner sleeve below the insert tube adjacent the top. The insert tube is relatively non-radioactivated inconel alloy. The opposed springs and dimples are formed on diametrically opposite inner walls of the insert tube, the springs being formed as spaced axial cuts in the insert tube, with a web of the insert tube between the cuts bowed radially inwardly for forming the spring, and the dimples being formed as radially inward protrusions opposed to the springs. 7 figures.

  11. Physical aspects of liquid-impelled loop reactors

    NARCIS (Netherlands)

    Sonsbeek, van H.

    1992-01-01

    The liquid-impelled loop reactor (LLR) is a reactor that consists of two parts : the main tube and the circulation tube. Both parts are in open connection at the bottom and at the top. The reactor is filled with a liquid phase: the continuous phase. Another liquid phase is injected in the

  12. 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

  13. 76 FR 60358 - Gypsy Moth Generally Infested Areas; Additions in Indiana, Maine, Ohio, Virginia, West Virginia...

    Science.gov (United States)

    2011-09-29

    .... APHIS-2010-0075] Gypsy Moth Generally Infested Areas; Additions in Indiana, Maine, Ohio, Virginia, West Virginia, and Wisconsin AGENCY: Animal and Plant Health Inspection Service, USDA. ACTION: Affirmation of... amended the regulations to add areas in Indiana, Maine, Ohio, Virginia, West Virginia, and Wisconsin to...

  14. 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 ...

  15. The future of transit in West Virginia.

    Science.gov (United States)

    2013-12-01

    The Future of Transit in West Virginia is a study of the current system of public transportation in West Virginia and : an examination of issues, priorities and projections of the public transportation network in the coming years. The : purpose...

  16. An approach to pavement management in Virginia.

    Science.gov (United States)

    1981-01-01

    The report summarizes the objectives and benefits of formal pavement management systems and outlines an approach believed by the author to be practical for Virginia. The management of Virginia interstate pavements and a proposed random-sampling plan ...

  17. Participation in the United States Department of Energy Reactor Sharing Program. Annual report, September 1981-August 1982

    International Nuclear Information System (INIS)

    Brenizer, J.S.; Benneche, P.E.

    1982-12-01

    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

  18. Safety issues at the defense production reactors

    International Nuclear Information System (INIS)

    1987-01-01

    The United States produces plutonium and tritium for use in nuclear weapons at the defense production reactors - the N Reactor in Washington and the Savannah River reactors in South Carolina. This report reaches general conclusions about the management of those reactors and highlights a number of safety and technical issues that should be resolved. The report provides an assessment of the safety management, safety review, and safety methodology employed by the Department of Energy and the private contractors who operate the reactors for the federal government. This report examines the safety objective established by the Department of Energy for the production reactors and the process the Department of its contractors use to implement the objective; focuses on a variety of uncertainties concerning the production reactors, particularly those related to potential vulnerabilities to severe accidents; and identifies ways in which the DOE approach to management of the safety of the production reactors can be improved

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Shirakawa, Toshihisa.

    1979-01-01

    Purpose: To prevent cladding tube injuries due to thermal expansion of each of the pellets by successively extracting each of the control rods loaded in the reactor core from those having less number of notches, as well as facilitate the handling work for the control rods. Constitution: A recycle flow control device is provided to a circulation pump for forcibly circulating coolants in the reactor container and an operational device is provided for receiving each of the signals concerning number of notches for each of the control rods and flow control depending on the xenon poisoning effect obtained from the signals derived from the in-core instrument system connected to the reactor core. The operational device is connected with a control rod drive for moving each of the control rods up and down and a recycle flow control device. The operational device is set with a pattern for the aimed control rod power and the sequence of extraction. Upon extraction of the control rods, they are extracted successively from those having less notch numbers. (Moriyama, K.)

  20. 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.

  1. Impact of Virginia's School-Entry Vaccine Mandate on Human Papillomavirus Vaccination Among 13-17-Year-Old Females.

    Science.gov (United States)

    Pierre-Victor, Dudith; Page, Timothy F; Trepka, Mary Jo; Stephens, Dionne P; Li, Tan; Madhivanan, Purnima

    2017-03-01

    The link between human papillomavirus (HPV) and anogenital cancers is well established in the literature. Many states have passed laws requiring funding for HPV education or vaccination. Mandatory HPV vaccination policies have been considered and passed in several states; yet their effectiveness has not been evaluated. This study sought to assess the impact of Virginia's HPV vaccine mandate for school-entry on HPV vaccine uptake among females aged 13-17 years. Data from the National Immunization Survey-Teen for the 2008-2012 period were used, and 3,203 adolescent females were included in the analysis. We performed difference-in-differences estimation and logistic regression with a policy and period interaction term. Virginia was considered the treatment state, and South Carolina and Tennessee were the comparison states to account for nonpolicy factors that may have affected vaccination rates during the time period considered in the analysis. There was no evidence of an effect of Virginia's HPV vaccine mandate for school-entry on vaccination rates or on physician vaccination recommendation using either the difference-by-differences analysis or the policy and period interaction term in the logistic regression. Physician recommendation was the factor most strongly associated with vaccination in the Virginia-South Carolina analysis (adjusted odds ratio [aOR] = 9.33; 95% confidence interval [CI]: 6.11-14.3) and in the Virginia-Tennessee analysis (aOR = 9.33; 95% CI: 6.11-14.3). Study findings suggest that Virginia's HPV vaccine mandate for school-entry did not lead to a significant increase in HPV vaccination among adolescent females or physician recommendations. However, physician recommendation was the factor most strongly associated with vaccination.

  2. History of research reactor fuel fabrication at Babcock and Wilcox

    International Nuclear Information System (INIS)

    Freim, James B.

    1983-01-01

    B and W Research Reactor Fuel Element facility at Lynchburg, Virginia now produces national laboratory and university fuel assemblies. The Company's 201000 square foot facility is devoted entirely to supplying research fuel and related products. B and W re-entered the research reactor fuel market in 1981

  3. Nuclear reactor containing facility

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Murase, Michio.

    1994-01-01

    In a reactor containing facility, a condensation means is disposed above the water level of a cooling water pool to condensate steams of the cooling water pool, and return the condensated water to the cooling water pool. Upon occurrence of a pipeline rupture accident, steams generated by after-heat of a reactor core are caused to flow into a bent tube, blown from the exit of the bent tube into a suppression pool and condensated in a suppression pool water, thereby suppressing the pressure in the reactor container. Cooling water in the cooling water pool is boiled by heat conduction due to the condensation of steams, then the steams are exhausted to the outside of the reactor container to remove the heat of the reactor container to the outside of the reactor. In addition, since cooling water is supplied to the cooling water pool quasi-permanently by gravity as a natural force, the reactor container can be cooled by the cooling water pool for a long period of time. Since the condensation means is constituted with a closed loop and interrupted from the outside, radioactive materials are never released to the outside. (N.H.)

  4. NCSU reactor sharing program. Final technical report

    International Nuclear Information System (INIS)

    Perez, P.B.

    1997-01-01

    The Nuclear Reactor Program at North Carolina State University provides the PULSTAR Research Reactor and associated facilities to eligible institutions with support, in part, from the Department of Energy Reactor Sharing Program. Participation in the NCSU Reactor Sharing Program continues to increase steadily with visitors ranging from advance high school physics and chemistry students to Ph.D. level research from neighboring universities. This report is the Final Technical Report for the DOE award reference number DE-FG05-95NE38136 which covers the period September 30, 1995 through September 30, 1996

  5. Neutron tubes

    Science.gov (United States)

    Leung, Ka-Ngo [Hercules, CA; Lou, Tak Pui [Berkeley, CA; Reijonen, Jani [Oakland, CA

    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.

  6. Iris reactor conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Carelli, M.D.; Conway, L.E.; Petrovic, B.; Paramonov, D.V. [Westinghouse Electric Comp., Pittsburgh, PA (United States); Galvin, M.; Todreas, N.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States); Lombardi, C.V.; Maldari, F.; Ricotti, M.E. [Politecnico di Milano, Milan (Italy); Cinotti, L. [Ansaldo SpA, Genoa (Italy)

    2001-07-01

    IRIS (International Reactor Innovative and Secure) is a modular, integral, light water cooled, low-to-medium power (100-350 MWe) reactor which addresses the requirements defined by the US DOE for Generation IV reactors, i.e., proliferation resistance, enhanced safety, improved economics and fuel cycle sustainability. It relies on the proven technology of light water reactors and features innovative engineering, but it does not require new technology development. This paper discusses the current reference IRIS design, which features a 1000 MWt thermal core with proven 5%-enriched uranium oxide fuel and five-year long straight burn fuel cycle, integral reactor vessel housing helical tube steam generators and immersed spool pumps. Other major contributors to the high level of safety and economic attractiveness are the safety by design and optimized maintenance approaches, which allow elimination of some classes of accidents, lower capital cost, long operating cycle, and high capacity factors. (author)

  7. 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...

  8. CALANDRIA TYPE SODIUM GRAPHITE REACTOR

    Science.gov (United States)

    Peterson, R.M.; Mahlmeister, J.E.; Vaughn, N.E.; Sanders, W.J.; Williams, A.C.

    1964-02-11

    A sodium graphite power reactor in which the unclad graphite moderator and fuel elements are contained within a core tank is described. The core tank is submersed in sodium within the reactor vessel. Extending longitudinally through the core thnk are process tubes with fuel elements positioned therein. A bellows sealing means allows axial expansion and construction of the tubes. Within the core tank, a leakage plenum is located below the graphite, and above the graphite is a gas space. A vent line regulates the gas pressure in the space, and another line removes sodium from the plenum. The sodium coolant flows from the lower reactor vessel through the annular space between the fuel elements and process tubes and out into the reactor vessel space above the core tank. From there, the heated coolant is drawn off through an outlet line and sent to the heat exchange. (AEC)

  9. Modifying the Heysham 2 and Torness guide tubes

    International Nuclear Information System (INIS)

    Salter, I.D.

    1988-01-01

    In 1986 the National Nuclear Corporation carried out the unfuelled engineering run on Torness Reactor One. Subsequent inspection revealed wear on the reactor control rods, following severe spinning caused by gas cross flow swirls at the guide tube castellated ring. The adopted solution was to machine 32 radial holes through the guide tube wall and blank off the existing castellated slots, however, man access to the guide tubes is extremely difficult. This paper describes how Taylor Hitec produced, in only 12 weeks, three remote drilling machines, together with associated debris collection systems, cleaning equipment and remote video/CCTV inspection systems, and then carried out the modifications to the reactors. (author)

  10. Control system for a small fission reactor

    Science.gov (United States)

    Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

    1985-02-08

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

  11. Manufacturing and testing the HTGR refueling tube

    International Nuclear Information System (INIS)

    Gurin, V.A.; Gribanov, Yu.A.; Kolosenko, V.V.; Gujda, V.V.

    2015-01-01

    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 1·10 21 n/cm 2 are comparable and meet the specifications for HTGR refueling tube

  12. FBR type reactor

    International Nuclear Information System (INIS)

    Nagai, Fumio.

    1979-01-01

    Purpose: To unify the temperature distribution in a nuclear reactor vessel by the provision of a gas recycle path for pressurizing a cover gas to recycle the cover gas and thus stir the gas in a cover gas chamber. Constitution: A plurality of gas inlet tubes and gas discharge tubes are provided to the wall of a cover gas chamber above the liquid level of coolants in a nuclear reactor vessel and the cover gas is recycled through the tubes. The plurality of gas inlet tubes are each provided at their tops with nozzles opening circumferentially and communicated to the outlet of a compressor. While on the other hand, the plurality of gas discharge tubes are communicated to the inlet of a compressor. Upon operation of the compressor, the pressurized cover gas is jetted out from the nozzles, swirls along the inner circumferential surface of the vessel and interrupts and stirs the vertical thermal convection. The gas, after swirling one-half of the inner circumferential surface of the vessel, automatically flows out of the gas discharging tubes opening behind the nozzles and then flows into the inlet of the compressor. (Seki, T.)

  13. Residential Energy Efficiency Potential: Virginia

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Eric J [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-11-27

    Energy used by Virginia single-family homes that can be saved through cost-effective improvements. Prepared by Eric Wilson and Noel Merket, NREL, and Erin Boyd, U.S. Department of Energy Office of Energy Policy and Systems Analysis.

  14. Fuel assembly in a reactor

    International Nuclear Information System (INIS)

    Saito, Shozo; Kawahara, Akira.

    1975-01-01

    Object: To provide a fuel assembly in a reactor which can effectively prevent damage of the clad tube caused by mutual interference between pellets and the clad tube. Structure: A clad tube for a fuel element, which is located in the outer peripheral portion, among the fuel elements constituting fuel assemblies arranged in assembled and lattice fashion within a channel box, is increased in thickness by reducing the inside diameter thereof to be smaller than that of fuel elements internally located, thereby preventing damage of the clad tube resulting from rapid rise in output produced when control rods are removed. (Kamimura, M.)

  15. 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.

  16. 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.

  17. 7 CFR 51.2753 - U.S. Virginia Splits.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 2 2010-01-01 2010-01-01 false U.S. Virginia Splits. 51.2753 Section 51.2753... STANDARDS) United States Standards for Shelled Virginia Type Peanuts Grades § 51.2753 U.S. Virginia Splits. “U.S. Virginia Splits” consists of shelled Virginia type peanut kernels of similar varietal...

  18. SQUIRT, Seepage in Reactor Tube Cracks

    International Nuclear Information System (INIS)

    Paul, D.; Ghadiali, N.; Wilkowski, G.; Rahman, S.; Krishnaswamy, P.

    1997-01-01

    1 - Description of program or function: The SQUIRT software is designed to perform leakage rate and area of crack opening calculations for through-wall cracks in pipes. The fluid in the piping system is assumed to be water at either subcooled or saturated conditions. The development of the SQUIRT computer model enables licensing authorities and industry users to conduct the leak-rate evaluations for leak-before-break applications in a more efficient manner. 2 - Method of solution: The SQUIRT program uses a modified form of the Henry-Fauske model for the thermal-hydraulics analysis together with Elastic-Plastic Fracture Mechanics using GE/EPRI and LBB.ENG2 methods for crack opening analysis. 3 - Restrictions on the complexity of the problem: Squirt requires 512 KB of conventional memory and an organized structure. Software can only be executed from the main SQUIRT23 directory where the software was installed

  19. Fuel assembly for pressure tube type reactor

    International Nuclear Information System (INIS)

    Wakabayashi, Toshio.

    1990-01-01

    For providing a effect of burnable poisons without worsening local power peaking, it is effective to dispose conventional fuel rods incorporated with burnable poisons uniformly to the outermost layer of a fuel assembly in which fuel rods are disposed concentrically in a multilayered state. However, since the irradiation is applied to the outermost layer with high neutron fluxes in this case, the burnable poisons are eliminated rapidly and excess reactivity suppression effect does not last sufficiently. Then, according to the present invention, pellets of a dual layer structure comprising a pellet containing burnable poisons disposed at the center and nuclear fuel materials composed of plutonium-uranium mixed oxides or uranium oxides coated therearound are disposed. In view of the above, it is possible to obtain a fuel assembly which does not increase the suppression of excess reactivity at the initial stage and does not promote the elimination of burnable poisons, without lowering the local power peaking. (T.M.)

  20. Nuclear reactor fuelling machine

    International Nuclear Information System (INIS)

    Peberdy, J.M.

    1976-01-01

    The refuelling machine described comprises a rotatable support structure having a guide tube attached to it by a parellel linkage mechanism, whereby the guide tube can be displaced sideways from the support structure. A gripper unit is housed within the guide tube for gripping the end of a fuel assembly or other reactor component and has means for maintenance in the engaging condition during travel of the unit along the guide tube, except for a small portion of the travel at one end of the guide tube, where the inner surface of the guide tube is shaped so as to maintain the gripper unit in a disengaging condition. The gripper unit has a rotatable head, means for moving it linearly within the guide tube so that a component carried by the unit can be housed in the guide tube, and means for rotating the head of the unit through 180 0 relative to its body, to effect rotation of a component carried by the unit. The means for rotating the head of the gripper unit comprises ring and pinion gearing, operable through a series of rotatable shafts interconnected by universal couplings. The reason for provision for 180 0 rotation is that due to the variation in the neutron flux across the reactor core the side of a fuel assembly towards the outside of the core will be subjected to a lower neutron flux and therefore will grow less than the side of the fuel assembly towards the inside of the core. This can lead to bowing and possible jamming of the fuel assemblies. Full constructional details are given. See also BP 1112384. (U.K.)

  1. Consequences of pressure tube rupture on in-core components

    International Nuclear Information System (INIS)

    Hill, P.G.; Hauptmann, E.G.; Lee, V.

    1982-12-01

    An investigation has been made of the consequences of pressure tube rupture in calandria vessels of heavy water cooled and moderated reactors. The study included a review of previous experimental and analytical work, as well as supplementary investigations carried out to examine the validity of previous assumptions and findings. The central questions considered were: the possibility of a propagating pressure tube failure; damage to the calandria vessel; and damage to the shut-off-rod guide tubes of the reactor shut-down system. The results of the investigation do not indicate mechanisms of sufficient strength to cause propagating failure in a well-designed, well-operated reactor following a tube burst under normal operating conditions. However, not all the details of the physical processes involved in a tube burst have been revealed by existing experimental and analytical work

  2. The effect of tube-support interaction on the dynamic response of heat exchanger tubes

    International Nuclear Information System (INIS)

    Shin, Y.S.; Jendrzejczyk, J.A.; Wambsganss, M.W.

    1977-01-01

    To avoid detrimental tube vibration in heat exchangers, resonant conditions and instabilitites must be avoided, and/or peak dynamic amplitudes must not exceed allowable limits. In attempting a theoretical analysis, questions arise as to the effects of tube/support interaction on tube vibrational characteristics (i.e. resonant frequencies, modes, damping) and response amplitude. As a part of ANL's Flow-Induced Vibration Program in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design activity, tube/support interaction experiments are being performed not only to gain the insight into the dynamic behavior of CRBRP steam generator tubes, but also to provide the basis for developing design guidance. Test results were compared with anaytical results based on multispan tube with 'knife-edge' supports at the support locations. (Auth.)

  3. Selected reading on introduction to pressure tube technology

    International Nuclear Information System (INIS)

    Causey, A.R.; Coleman, C.E.; Ells, C.E.

    1981-10-01

    Four lectures on pressure tube technology were presented at Sheridan Park, Ontario, on 1981 June 1. The titles were 'Pressure Tubes and Their Operational Environment', 'Fabrication, Inspection and Properties of Current Production Pressure Tubes', 'In-Reactor Deformation of Fuel Channels', and 'Potential Failure Modes in Pressure Tubes'. This report lists the references used in preparing the lectures. It is intended to provide a starting point in reading for people who need to become familiar with pressure tube technology but have little prior knowledge of the topic

  4. 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.

  5. Virginia Power's regulatory reduction program

    International Nuclear Information System (INIS)

    Miller, G.D.

    1996-01-01

    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

  6. South Carolina Kids Count, 2001.

    Science.gov (United States)

    Holmes, A. Baron

    This Kids Count report examines statewide trends in the well-being of South Carolina's children. The statistical portrait is based on 42 indicators in the areas of demographics, family, economic status, health, readiness and early school performance, scholastic achievement, and adolescent risk behaviors. The indicators are: (1) population; (2)…

  7. South Carolina Kids Count, 2000.

    Science.gov (United States)

    Holmes, A. Baron

    This Kids Count report examines statewide trends in the well-being of South Carolina's children. The statistical portrait is based on 41 indicators in the areas of demographics, family, economic status, health, readiness and early school performance, scholastic achievement, and adolescent risk behaviors. The indicators are: (1) population; (2)…

  8. Ceramic oxygen transport membrane array reactor and reforming method

    Science.gov (United States)

    Kelly, Sean M.; Christie, Gervase Maxwell; Robinson, Charles; Wilson, Jamie R; Gonzalez, Javier E.; Doraswami, Uttam R.

    2017-10-03

    The invention relates to a commercially viable modular ceramic oxygen transport membrane system for utilizing heat generated in reactively-driven oxygen transport membrane tubes to generate steam, heat process fluid and/or provide energy to carry out endothermic chemical reactions. The system provides for improved thermal coupling of oxygen transport membrane tubes to steam generation tubes or process heater tubes or reactor tubes for efficient and effective radiant heat transfer.

  9. Virginia Apgar and string music

    OpenAIRE

    Palacios-Sánchez, Leonardo

    2011-01-01

    Virginia Apgar (1909-1974) is one of the most recognized American doctors, worldwide known by his contribution as the developer of the "Apgar test" a method used for the evaluation of newborns all over the world. She had many interests. She was anesthesiologist, a brilliant teacher and researcher, but she also loved lecture, basketball, fishing, golf, philately, and music. She played violin and cello and she interpreted that instruments in various chamber groups. Being motivated by one of her...

  10. Manumission in Nineteenth Century Virginia

    OpenAIRE

    Howard Bodenhorn

    2010-01-01

    A long-standing debate concerns the rationality of slave owners and this paper addresses that debate within the context of manumission. Using a new sample of 19th-century Virginia manumissions, I show that manumission was associated with the productive characteristics of slaves. More productive slaves were manumitted at younger ages than less productive slaves. Although more productive slaves were more valuable to slave owners, which might be expected to delay manumission, more productive sla...

  11. Virginia Power's nuclear operations: Leading by example

    International Nuclear Information System (INIS)

    Kuehn, S.E.

    1995-01-01

    Success has been a long time coming for Virginia Power's nuclear units, but after a record run and some of the shortest refueling outages ever, the rest of the industry could learn a few things. This article describes the changes made by Virginia Power at its Surry and North Anna plants. Virginia Power's recipe for success called for equal amounts of individual initiative, management savvy, engineering discipline, organization, dedication, perseverance, pride, introspection, motivation, and humility

  12. SSN 774 Virginia Class Submarine (SSN 774)

    Science.gov (United States)

    2015-12-01

    Selected Acquisition Report (SAR) RCS: DD-A&T(Q&A)823-516 SSN 774 Virginia Class Submarine (SSN 774) As of FY 2017 President’s Budget Defense...December 2015 SAR March 8, 2016 11:22:44 UNCLASSIFIED 4 CAPT Michael Stevens VIRGINIA Submarine Program Office PEO Submarines 614 Sicard Street, SE...16, 2015 Program Information Program Name SSN 774 Virginia Class Submarine (SSN 774) DoD Component Navy Responsible Office References SAR

  13. 77 FR 27120 - Safety Zone; Virginia Beach Oceanfront Air Show, Atlantic Ocean, Virginia Beach, VA

    Science.gov (United States)

    2012-05-09

    ... SECURITY Coast Guard 33 CFR Part 165 RIN 1625-AA00 Safety Zone; Virginia Beach Oceanfront Air Show, Atlantic Ocean, Virginia Beach, VA AGENCY: Coast Guard, DHS. ACTION: Temporary final rule. SUMMARY: The... Beach, VA to support the Virginia Beach Oceanfront Air Show. This action is necessary to provide for the...

  14. 77 FR 13519 - Safety Zone; Virginia Beach Oceanfront Air Show, Atlantic Ocean, Virginia Beach, VA

    Science.gov (United States)

    2012-03-07

    ... SECURITY Coast Guard 33 CFR Part 165 RIN 1625-AA00 Safety Zone; Virginia Beach Oceanfront Air Show, Atlantic Ocean, Virginia Beach, VA AGENCY: Coast Guard, DHS. ACTION: Notice of proposed rulemaking. SUMMARY... Virginia Beach, VA. This action is necessary to provide for the safety of life on navigable waters during...

  15. 76 FR 54189 - Television Broadcasting Services; Hampton-Norfolk, Virginia; Norfolk, Virginia-Elizabeth City, NC

    Science.gov (United States)

    2011-08-31

    ...] Television Broadcasting Services; Hampton-Norfolk, Virginia; Norfolk, Virginia-Elizabeth City, NC AGENCY... licensee of noncommercial educational television station WHRO-TV, channel *16, Hampton-Norfolk, Virginia... freeze on the filing of television allotment rulemaking petitions, but since HRETA'S proposal...

  16. 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.

  17. Validation of NESTLE against static reactor benchmark problems

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1996-01-01

    The NESTLE advanced modal code was developed at North Carolina State University with support from Los Alamos National Laboratory and Idaho National Engineering Laboratory. It recently has been benchmarked successfully against measured data from pressurized water reactors (PWRs). However, NESTLE's geometric capabilities are very flexible, and it can be applied to a variety of other types of reactors. This study presents comparisons of NESTLE results with those from other codes for static benchmark problems for PWRs, boiling water reactors (BWRs), high-temperature gas-cooled reactors (HTGRs) and CANDU heavy- water reactors (HWRs)

  18. Validation of NESTLE against static reactor benchmark problems

    International Nuclear Information System (INIS)

    Mosteller, R.D.

    1996-01-01

    The NESTLE advanced nodal code was developed at North Carolina State University with support from Los Alamos National Laboratory and Idaho National Engineering Laboratory. It recently has been benchmarked successfully against measured data from pressurized water reactors (PWRs). However, NESTLE's geometric capabilities are very flexible, and it can be applied to a variety of other types of reactors. This study presents comparisons of NESTLE results with those from other codes for static benchmark problems for PWRs, boiling water reactors (BWRs), high-temperature gas-cooled reactors (HTGRs), and Canada deuterium uranium (CANDU) heavy-water reactors (HWRs)

  19. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  20. Virginia Water Center provides online inventory of water-related bills in the 2009 Virginia General Assembly

    OpenAIRE

    Fay, Patrick

    2009-01-01

    The Virginia Water Resources Research Center at Virginia Tech has released a list of over 170 water resource-related bills in the 2009 Virginia General Assembly, and has made it available at its website.

  1. Automatic integrated testing bench for tubes in translation

    International Nuclear Information System (INIS)

    Dufayet, J.P.; Perdijon, J.

    1976-01-01

    All the nondestructive tests required for receiving the cladding tubes intended for fast nuclear reactor are integrated on this bench: quality control by eddy currents and ultra-sounds, thickness and (inner and outer) diameter measurement. The linear displacement of the tube allows very high rates to be attained [fr

  2. Tracheostomy tube - eating

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/patientinstructions/000464.htm Tracheostomy tube - eating To use the sharing features on ... when you swallow foods or liquids. Eating and Tracheostomy Tubes When you get your tracheostomy tube, or ...

  3. Eustachian tube patency

    Science.gov (United States)

    Eustachian tube patency refers to how much the eustachian tube is open. The eustachian tube runs between the middle ear and the throat. It controls the pressure behind the eardrum and middle ear space. This helps keep ...

  4. Multifrequency eddy current testing of helical tubes of steam generators

    International Nuclear Information System (INIS)

    Pigeon, M.; David, B.

    1983-06-01

    In the event of a water-sodium reaction in a steam-generator of a fast breeder reactor, it is necessary to test the tubes close to the leak to evaluate the damage. In SUPERPHENIX, the tubes are about 100m long and are coiled on a dead body. This report describes the equipment and the technic to test such tubes with multifrequency eddy current technics [fr

  5. A method and machine for forming pleated and bellow tubes

    International Nuclear Information System (INIS)

    Banks, J.W.

    1975-01-01

    In a machine, the rollers outside the rough tube are rigidly supported for assuring the accurate forming of each turn of the pleated tube, the latter being position-indexed independently of the already formed turns. An inner roller is supported by a device for adjusting and indexing the position thereof on a carriage. The thus obtained tubes are suitable, in particular, for forming expansion sealing joints for power generators or nuclear reactors [fr

  6. Reactor water sampling device

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo.

    1992-01-01

    The present invention concerns a reactor water sampling device for sampling reactor water in an in-core monitor (neutron measuring tube) housing in a BWR type reactor. The upper end portion of a drain pipe of the reactor water sampling device is attached detachably to an in-core monitor flange. A push-up rod is inserted in the drain pipe vertically movably. A sampling vessel and a vacuum pump are connected to the lower end of the drain pipe. A vacuum pump is operated to depressurize the inside of the device and move the push-up rod upwardly. Reactor water in the in-core monitor housing flows between the drain pipe and the push-up rod and flows into the sampling vessel. With such a constitution, reactor water in the in-core monitor housing can be sampled rapidly with neither opening the lid of the reactor pressure vessel nor being in contact with air. Accordingly, operator's exposure dose can be reduced. (I.N.)

  7. Reactor Safety Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  8. Heat exchanger tube mounts

    Science.gov (United States)

    Wolowodiuk, W.; Anelli, J.; Dawson, B.E.

    1974-01-01

    A heat exchanger in which tubes are secured to a tube sheet by internal bore welding is described. The tubes may be moved into place in preparation for welding with comparatively little trouble. A number of segmented tube support plates are provided which allow a considerable portion of each of the tubes to be moved laterally after the end thereof has been positioned in preparation for internal bore welding to the tube sheet. (auth)

  9. Nuclear reactor machine refuelling system

    International Nuclear Information System (INIS)

    Cashen, W.S.; Erwin, D.

    1977-01-01

    Part of an on-line fuelling machine for a CANDU pressure-tube reactor is described. The present invention provides a refuelling machine wherein the fuelling components, including the fuel carrier and the closure adapter, are positively positioned and retained within the machine magazine or positively secured to the machine charge tube head, and cannot be accidentally disengaged as in former practice. The positive positioning devices include an arcuate keeper plate. Simplified hooked fingers are used. (NDH)

  10. Tube holding system

    International Nuclear Information System (INIS)

    Cunningham, R.C.

    1978-01-01

    A tube holding rig is described for the lateral support of tubes arranged in tight parcels in a heat exchanger. This tube holding rig includes not less than two tube supporting assemblies, with a space between them, located crosswise with respect to the tubes, each supporting assembly comprising a first set of parallel components in contact with the tubes, whilst a second set of components is also in contact with the tubes. These two sets of parts together define apertures through which the tubes pass [fr

  11. 77 FR 2766 - Facility Operating License Amendment from Duke Energy Carolinas, LLC., Catawba Nuclear Station...

    Science.gov (United States)

    2012-01-19

    ..., Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory... from periodic SG tube inspections and plugging, Permanently reduce the primary to secondary leakage... not affected by the primary to secondary leakage flow during the event, as primary to secondary...

  12. Scanning tunneling microscope assembly, reactor, and system

    Science.gov (United States)

    Tao, Feng; Salmeron, Miquel; Somorjai, Gabor A

    2014-11-18

    An embodiment of a scanning tunneling microscope (STM) reactor includes a pressure vessel, an STM assembly, and three spring coupling objects. The pressure vessel includes a sealable port, an interior, and an exterior. An embodiment of an STM system includes a vacuum chamber, an STM reactor, and three springs. The three springs couple the STM reactor to the vacuum chamber and are operable to suspend the scanning tunneling microscope reactor within the interior of the vacuum chamber during operation of the STM reactor. An embodiment of an STM assembly includes a coarse displacement arrangement, a piezoelectric fine displacement scanning tube coupled to the coarse displacement arrangement, and a receiver. The piezoelectric fine displacement scanning tube is coupled to the coarse displacement arrangement. The receiver is coupled to the piezoelectric scanning tube and is operable to receive a tip holder, and the tip holder is operable to receive a tip.

  13. Development of Zirconium alloys (for pressure tubes)

    International Nuclear Information System (INIS)

    Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Jung, Chung Hwan; Yim, Kyong Soo; Kim, Sung Soo; Baek, Jong Hyuk; Jeong, Yong Hwan; Kim, Kyong Ho; Cho, Hae Dong; Hwang, S. K.; Kim, M. H.; Kwon, S. I; Kim, I. S.

    1997-09-01

    The objective of this research is to set up the basic technologies for the evaluation of pressure tube integrity and to develop improved zirconium alloys to prevent pressure tube failures due to DHC and hydride blister caused by excessive creep-down of pressure tubes. The experimental procedure and facilities for characterization of pressure tubes were developed. The basic research related to a better understanding of the in-reactor performances of pressure tubes leads to noticeable findings for the first time : the microstructural effect on corrosion and hydrogen pick-up behavior of Zr-2.5Nb pressure tubes, texture effect on strength and DHC resistance and enhanced recrystallization by Fe in zirconium alloys and etc. Analytical methodology for the assessment of pressure tubes with surface flaws was set up. A joint research is being under way with AECL to determine the fracture toughness of O-8 at the EOL (End of Life) that had been quadruple melted and was taken out of the Wolsung Unit-1 after 10 year operation. In addition, pressure tube with texture controlled is being made along with VNINM in Russia as a joint project between KAERI and Russia. Finally, we succeeded in developing 4 different kinds of zirconium alloys with better corrosion resistance, low hydrogen pickup fraction and higher creep strength. (author). 121 refs., 65 tabs., 260 figs

  14. Steam generator tube integrity program

    International Nuclear Information System (INIS)

    Diercks, D.R.; Shack, W.J.

    1996-01-01

    The degradation of steam generator tubes in pressurized water nuclear reactors continues to be a serious problem, and the US Nuclear Regulatory Commission (NRC) is developing a performance-based rule and regulatory guide for steam generator tube integrity. To support the evaluation of industry-proposed implementation of these performance-based criteria, the NRC is sponsoring a new research program at Argonne National Laboratory on steam generator tubing degradation. The objective of the new program is to provide the necessary experimental data and predictive correlations and models that will permit the NRC to independently evaluate the integrity of steam generator tubes. The technical work in the program is divided into four tasks, (1) assessment of inspection reliability, (2) research on in-service inspection technology, (3) research on degradation modes and integrity, and (4) development of methodology and technical assessments for current and emerging regulatory issues. The objectives of and planned research activities under each of these four tasks are described here. (orig.)

  15. 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.

  16. Virginia Agribusiness Council honors Veterinarian Whittier

    OpenAIRE

    Douglas, Jeffrey S.

    2006-01-01

    Years of hard work by one of the state's most prominent university-based extension agents were recently recognized when the Virginia Agribusiness Council awarded Virginia Tech's Production Management Medicine/Bovine Specialist - Veterinary Extension Dr. W. D. Whittier their annual"Extension Service Award."

  17. Forests of Virginia, 2014

    Science.gov (United States)

    Anita Rose

    2016-01-01

    This resource update provides an overview of forest resources in Virginia. Information for this factsheet was updated by means of the Forest Inventory and Analysis (FIA) annualized sample design. Each year, 20 percent of the sample plots (one panel) in Virginia are measured by field crews, the data compiled, and new estimates produced. After 5 years of measurements,...

  18. 40 CFR 81.347 - Virginia.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 17 2010-07-01 2010-07-01 false Virginia. 81.347 Section 81.347... AREAS FOR AIR QUALITY PLANNING PURPOSES Section 107 Attainment Status Designations § 81.347 Virginia. Virginia—TSP Designated area Does not meet primary standards Does not meet secondary standards Cannot be...

  19. 40 CFR 81.433 - Virginia.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 17 2010-07-01 2010-07-01 false Virginia. 81.433 Section 81.433 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) DESIGNATION OF... Visibility Is an Important Value § 81.433 Virginia. Area name Acreage Public Law establishing Federal land...

  20. Virginia Standards Predated the Common Core Initiative

    Science.gov (United States)

    Knowledge Quest, 2014

    2014-01-01

    The Virginia Board of Education is committed to the Virginia Standards of Learning (SOL) program and opposed to adoption of the newly developed Common Core State Standards as a prerequisite for participation in federal competitive grant and entitlement programs. The Standards of Learning are clear and rigorous and have won the acceptance and trust…

  1. 40 CFR 81.349 - West Virginia.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 17 2010-07-01 2010-07-01 false West Virginia. 81.349 Section 81.349... Virginia. West Virginia—TSP Designated area Does not meet primary standards Does not meet secondary... district in Berkeley County X Remainder of State X West Virginia—SO2 Designated area Does not meet primary...

  2. 40 CFR 81.435 - West Virginia.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 17 2010-07-01 2010-07-01 false West Virginia. 81.435 Section 81.435 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) DESIGNATION OF... Visibility Is an Important Value § 81.435 West Virginia. Area name Acreage Public Law establishing Federal...

  3. Virginia's College and Career Readiness Initiative

    Science.gov (United States)

    Virginia Department of Education, 2010

    2010-01-01

    In 1995, Virginia began a broad educational reform program that resulted in revised, rigorous content standards, the Virginia Standards of Learning (SOL), in the content areas of English, mathematics, science, and history and social science. These grade-by-grade and course-based standards were developed over 14 months with revision teams including…

  4. Virginia, 2009 forest inventory and analysis factsheet

    Science.gov (United States)

    Anita K. Rose

    2011-01-01

    This science update is a brief look at some of the basic metrics that describe forest resources in Virginia. Estimates presented here are for the measurement year 2009. Information for the factsheet is updated by means of the Forest Inventory and Analysis (FIA) annualized sample design. Virginia has about 4,600 sample plots across the State, and each year 20 percent of...

  5. Virginia, 2010 forest inventory and analysis factsheet

    Science.gov (United States)

    Anita K. Rose

    2012-01-01

    This science update is a brief look at some of the basic metrics that describe the status of forest resources in Virginia. Estimates presented here are for the measurement year 2010. Information for this factsheet is updated by means of the Forest Inventory and Analysis (FIA) annualized sample design. Virginia has about 4,600 sample plots across the State and each year...

  6. Failure analysis of retired steam generator tubings

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hong Pyo; Kim, J. S.; Hwang, S. S. and others

    2005-04-15

    Degradation of steam generator leads to forced outage and extension of outage, which causes increase in repair cost, cost of purchasing replacement power and radiation exposure of workers. Steam generator tube rupture incident occurred in Uljin 4 in 2002, which made public sensitive to nuclear power plant. To keep nuclear energy as a main energy source, integrity of steam generator should be demonstrated. Quantitative relationship between ECT(eddy current test) signal and crack size is needed in assesment of integrity of steam generator in pressurized water reactor. However, it is not fully established for application in industry. Retired steam generator of Kori 1 has many kinds of crack such as circumferential and axial primary water stress corrosion crack and outer diameter stress corrosion crack(ODSCC). So, it can be used in qualifying and improving ECT technology and in condition monitoring assesment for crack detected in ISI(in service inspection). In addition, examination of pulled tube of Kori 1 retired steam generator will give information about effectiveness of non welded sleeving technology which was employed to repair defect tubes and remedial action which was applied to mitigate ODSCC. In this project, hardware such as semi hot lab. for pulled tube examination and modification transportation cask for pulled tube and software such as procedure of transportation of radioactive steam generator tube and non-destructive and destructive examination of pulled tube were established. Non-destructive and destructive examination of pulled tubes from Kori 1 retired steam generator were performed in semi hot lab. Remedial actions applied to Kori 1 retired steam generator, PWSCC trend and bulk water chemistry and crevice chemistry in Kori 1 were evaluated. Electrochemical decontamination technology for pulled tube was developed to reduce radiation exposure and enhance effectiveness of pulled tube examination. Multiparameter algorithm developed at ANL, USA was

  7. Reactor operation environmental information document

    Energy Technology Data Exchange (ETDEWEB)

    Wike, L.D.; Specht, W.L.; Mackey, H.E.; Paller, M.H.; Wilde, E.W.; Dicks, A.S.

    1989-12-01

    The Savannah River Site (SRS) is a large United States Department of Energy installation on the upper Atlantic Coastal Plain of South Carolina. The SRS contains diverse habitats, flora, and fauna. Habitats include upland terrestrial areas, varied wetlands including Carolina Bays, the Savannah River swamp system, and impoundment related and riparian wetlands, and the aquatic habitats of several stream systems, two large cooling reservoirs, and the Savannah River. These diverse habitats support a large variety of plants and animals including many commercially or recreational valuable species and several rare, threatened or endangered species. This volume describes the major habitats and their biota found on the SRS, and discuss the impacts of continued operation of the K, L, and P production reactors.

  8. Computation and measurement of calandria tube sag in PHWR

    International Nuclear Information System (INIS)

    Kim, Tae Ryong; Sohn, Seok Man

    2003-01-01

    Calandria tubes and liquid injection shutdown system (LISS) tubes in a pressurized heavy water reactor (PHWR) is known 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 LISS tube crossing beneath and 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 measuring probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sags of both tubes in the PHWR. 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. (author)

  9. H Reactor

    Data.gov (United States)

    Federal Laboratory Consortium — The H Reactor was the first reactor to be built at Hanford after World War II.It became operational in October of 1949, and represented the fourth nuclear reactor on...

  10. Nursing advocacy in North Carolina.

    Science.gov (United States)

    Gosselin-Acomb, Tracy K; Schneider, Susan M; Clough, Robert W; Veenstra, Brittney A

    2007-09-01

    To identify the ways oncology nurses in one state advocate for patients, as well as the resources they use to do so. Descriptive, cross-sectional survey. North Carolina. 141 RNs in North Carolina who were members of the Oncology Nursing Society (ONS). Subjects completed a two-page, self-administered questionnaire comprised of fixed-choice and open-ended questions. Demographics, frequency of advocating for patient services, and awareness of ONS resources. Nurses in North Carolina advocate for patients in a variety of ways. A need exists to develop ongoing methods to keep nurses up to date on advocacy issues, as well as to establish mentoring opportunities for them. Nurses believe that they are most challenged in addressing patients' financial and insurance concerns. Oncology nurses frequently advocate for patients' needs. The findings provide direction for future initiatives to educate nurses about their role in patient advocacy and available resources. Ongoing education and research are needed to enhance the role of oncology nurses as patient advocates.

  11. Compression device for feeding a waste material to a reactor

    Science.gov (United States)

    Williams, Paul M.; Faller, Kenneth M.; Bauer, Edward J.

    2001-08-21

    A compression device for feeding a waste material to a reactor includes a waste material feed assembly having a hopper, a supply tube and a compression tube. Each of the supply and compression tubes includes feed-inlet and feed-outlet ends. A feed-discharge valve assembly is located between the feed-outlet end of the compression tube and the reactor. A feed auger-screw extends axially in the supply tube between the feed-inlet and feed-outlet ends thereof. A compression auger-screw extends axially in the compression tube between the feed-inlet and feed-outlet ends thereof. The compression tube is sloped downwardly towards the reactor to drain fluid from the waste material to the reactor and is oriented at generally right angle to the supply tube such that the feed-outlet end of the supply tube is adjacent to the feed-inlet end of the compression tube. A programmable logic controller is provided for controlling the rotational speed of the feed and compression auger-screws for selectively varying the compression of the waste material and for overcoming jamming conditions within either the supply tube or the compression tube.

  12. Unsteady two dimensional multiphysical simulation on the radiating calandria tube under the subcooling boundary condition

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Se Myong [Kunsan National Univ., Kunsan (Korea, Republic of); Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Inside the Calandria tubes in the moderator system of a heavy water reactor, there are pressure tubes undergoing high pressure and temperature. If the cooling water dries out due to the local film boiling at the outer tube boundary, the excessive heat flux can deform the pressure tube to even contact with outer Calandria tube. To limit the subcooling for the avoidance of dryout condition in a CANDU reactor, a suitable experiment should be proposed such as Ref.. In this study, we simulated this experiment in 2 D with COMSOL Multi physics.

  13. STEAM STIRRED HOMOGENEOUS NUCLEAR REACTOR

    Science.gov (United States)

    Busey, H.M.

    1958-06-01

    A homogeneous nuclear reactor utilizing a selfcirculating liquid fuel is described. The reactor vessel is in the form of a vertically disposed tubular member having the lower end closed by the tube walls and the upper end closed by a removal fianged assembly. A spherical reaction shell is located in the lower end of the vessel and spaced from the inside walls. The reaction shell is perforated on its lower surface and is provided with a bundle of small-diameter tubes extending vertically upward from its top central portion. The reactor vessel is surrounded in the region of the reaction shell by a neutron reflector. The liquid fuel, which may be a solution of enriched uranyl sulfate in ordinary or heavy water, is mainiained at a level within the reactor vessel of approximately the top of the tubes. The heat of the reaction which is created in the critical region within the spherical reaction shell forms steam bubbles which more upwardly through the tubes. The upward movement of these bubbles results in the forcing of the liquid fuel out of the top of these tubes, from where the fuel passes downwardly in the space between the tubes and the vessel wall where it is cooled by heat exchangers. The fuel then re-enters the critical region in the reaction shell through the perforations in the bottom. The upper portion of the reactor vessel is provided with baffles to prevent the liquid fuel from splashing into this region which is also provided with a recombiner apparatus for recombining the radiolytically dissociated moderator vapor and a control means.

  14. Southeast Offshore Storage Resource Assessment (SOSRA): Evaluation of CO2 Storage Potential on the Continental Shelf from North Carolina to Florida

    Science.gov (United States)

    Knapp, J. H.; Knapp, C. C.; Brantley, D.; Lakshmi, V.; Howard, S.

    2016-12-01

    The Southeast Offshore Storage Resource Assessment (SOSRA) project is part of a major new program, funded by the U.S. Department of Energy for the next two and a half years, to evaluate the Atlantic and Gulf of Mexico offshore margins of the United States for geologic storage capacity of CO2. Collaborating organizations include the Southern States Energy Board, Virginia Polytechnic Institute, University of South Carolina, Oklahoma State University, Virginia Department of Mines, Minerals, and Energy, South Carolina Geological Survey, and Geological Survey of Alabama. Team members from South Carolina are focused on the Atlantic offshore, from North Carolina to Florida. Geologic sequestration of CO2 is a major research focus globally, and requires robust knowledge of the porosity and permeability distribution in upper crustal sediments. Using legacy seismic reflection, refraction, and well data from a previous phase of offshore petroleum exploration on the Atlantic margin, we are analyzing the rock physics characteristics of the offshore Mesozoic and Cenozoic stratigraphy on a regional scale from North Carolina to Florida. Major features of the margin include the Carolina Trough, the Southeast Georgia Embayment, the Blake Plateau basin, and the Blake Outer Ridge. Previous studies indicate sediment accumulations on this margin may be as thick as 12-15 km. The study will apply a diverse suite of data analysis techniques designed to meet the goal of predicting storage capacity to within ±30%. Synthetic seismograms and checkshot surveys will be used to tie well and seismic data. Seismic interpretation and geophysical log analysis will employ leading-edge software technology and state-of-the art techniques for stratigraphic and structural interpretation and the definition of storage units and their physical and chemical properties. This approach will result in a robust characterization of offshore CO2 storage opportunities, as well as a volumetric analysis that is

  15. Development of crystallographic texture in CANDU calandria tubes

    International Nuclear Information System (INIS)

    Theaker, J.R.; Coleman, C.E.

    2002-01-01

    The Zircaloy-2 calandria tubes in a CANDU nuclear reactor separate the hot Zr-2.5Nb pressure tubes from the cool moderator. These tubes are about 6 m long, have an outside diameter of 132 mm, and a wall thickness of 1.4 mm. To date, their performance has been exemplary. A possible feature for future reactors is to increase the strength of these calandria tubes to reduce the economic consequences of a hypothetical accident. The current method of fabrication is to form a sheet of Zircaloy-2 into a cylinder, then weld along the length. In fixed-end burst tests such tubes always fracture in the weld area because of the differences in crystallographic texture between the parent metal and the weld; eliminating the weld would increase the strength and ductility of the tube. We have evaluated four manufacturing routes for seamless tubes. To realize high biaxial strength, we require a large fraction of basal plane normals in the radial direction, F R . This paper describes these manufacturing routes, the calandria tube properties generated by the individual manufacturing routes, and their applicability for the CANDU system. The results show that the biaxial strength of a seamless calandria tube becomes greater with an increase in F R . which is related to the amount of cold work used to make the tubes, with saturation in F R after about 95% cold work. The results are interpreted in terms of anisotropic factors determined from uniaxial tension tests. (author)

  16. Reactor for Photocatalytic Degradation of Chloroform

    DEFF Research Database (Denmark)

    Simonsen, Morten Enggrob; Søgaard, Erik Gydesen

    In the present study a new type of continuous photoreactor is developed in which the TiO2 catalyst is immobilized on the surface of quartz tubes surrounding the UV lamps and on the internal surface of the reactor walls. The study showed that an initial concentration chloroform of 7 mg...... of the coated lamp in the reactor yield different degradation rates....

  17. 7 CFR 51.2751 - U.S. Medium Virginia.

    Science.gov (United States)

    2010-01-01

    ... 7 Agriculture 2 2010-01-01 2010-01-01 false U.S. Medium Virginia. 51.2751 Section 51.2751... STANDARDS) United States Standards for Shelled Virginia Type Peanuts Grades § 51.2751 U.S. Medium Virginia. “U.S. Medium Virginia” consists of shelled Virginia type peanut kernels of similar varietal...

  18. Assessment of the Impact of Viticulture Extension Programs in Virginia

    Science.gov (United States)

    Ferreira, Gustavo F. C.; Hatch, Tremain; Wolf, Tony K.

    2016-01-01

    The study discussed in this article assessed the impact of Virginia Cooperative Extension (VCE) on the Virginia wine grape industry. An online survey was developed and administered to members of the Virginia Vineyards Association. The results indicate that the resources and recommendations VCE and Virginia Tech have provided have been beneficial…

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Irion, L.; Tautz, J.; Ulrych, G.

    1976-01-01

    This additional patent complements the arrangement of non-return valves to prevent loss of cooling water on fracture of external tubes in the main coolant circuit (according to PS 24 24 427.7) by ensuring that the easily movable valves only operate in case of a fault, but do not flutter in operation, because the direction of flow is not the same at each location where they are installed. The remedy for this undesirable effect consists of allocating 1 non-return valve unit with 5 to 10 valves to each (of several) ducts for the cooling water intake. These units are installed in the annular space between the reactor vessel and the pressure vessel below the inlet of the ducts. Due to flow guidance surfaces in the same space, the incoming cooling water is deflected downwards and as the guiding surfaces are closed at the sides, must pass parallel to the valves of the non-return valve unit. On fracture of the external cooling water inlet pipe concerned, all valves of this unit close due to reversal of flow on the outlet side. (TK) [de

  20. Coastal storm monitoring in Virginia

    Science.gov (United States)

    Wicklein, Shaun M.; Bennett, Mark

    2014-01-01

    Coastal communities in Virginia are prone to flooding, particularly during hurricanes, nor’easters, and other coastal low-pressure systems. These weather systems affect public safety, personal and public property, and valuable infrastructure, such as transportation, water and sewer, and electric-supply networks. Local emergency managers, utility operators, and the public are tasked with making difficult decisions regarding evacuations, road closures, and post-storm recovery efforts as a result of coastal flooding. In coastal Virginia these decisions often are made on the basis of anecdotal knowledge from past events or predictions based on data from monitoring sites located far away from the affected area that may not reflect local conditions. Preventing flood hazards, such as hurricane-induced storm surge, from becoming human disasters requires an understanding of the relative risks that flooding poses to specific communities. The risk to life and property can be very high if decisions about evacuations and road closures are made too late or not at all.

  1. Nuclear reactor safety device

    Science.gov (United States)

    Hutter, Ernest

    1986-01-01

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  2. Rejection index for pressure tubes

    International Nuclear Information System (INIS)

    Mitchell, A.B.; Meneley, D.

    1989-10-01

    The objective of the present study was to establish a set of criteria (or Rejection Index) which could be used to decide whether a zirconium-2 1/2 w/o niobium pressure tube in a CANDU reactor should be removed from service due to in-service degradation. A critique of key issues associated with establishing a realistic rejection index was prepared. Areas of uncertainty in available information were identified and recommendations for further analysis and laboratory testing made. A Rejection Index based on the following limits has been recommended: 1) Limits related to design intent and normal operation: any garter spring must remain within the tolerance band specified for its design location; the annulus gas system must normally be operated in a circulating mode with a procedure in place for purging to prevent accumulation of deuterium. It must remain sensitive to leaks into any part of the systems; and pressure tube dimensions and distortions must be limited to maintain the fuel channels within the original design intent; 2) Limits related to defect tolerance: adequate time margins between occurrence of a leaking crack and unstable failure must be demonstrated for all fuel channels; long lap-type flaws are unacceptable; crack-like defects of any size are unacceptable; and score marks, frat marks and other defects with contoured profiles must fall below certain depth, length and stress intensity limits; and 3) Limits related to property degradation: at operating temperature each pressure tube must be demonstrated to have a critical length in excess of a stipulated value; the maximum equivalent hydrogen level in any pressure tube should not exceed a limit which should be defined taking into account the known history of that tube; the maximum equivalent hydrogen level in any rolled joint should not exceed a limit which is presently recommended as 200 ppm equivalent hydrogen; and the maximum diametral creep strain should be limited to less than 5%

  3. Oscillatory flow chemical reactors

    Directory of Open Access Journals (Sweden)

    Slavnić Danijela S.

    2014-01-01

    Full Text Available Global market competition, increase in energy and other production costs, demands for high quality products and reduction of waste are forcing pharmaceutical, fine chemicals and biochemical industries, to search for radical solutions. One of the most effective ways to improve the overall production (cost reduction and better control of reactions is a transition from batch to continuous processes. However, the reactions of interests for the mentioned industry sectors are often slow, thus continuous tubular reactors would be impractically long for flow regimes which provide sufficient heat and mass transfer and narrow residence time distribution. The oscillatory flow reactors (OFR are newer type of tube reactors which can offer solution by providing continuous operation with approximately plug flow pattern, low shear stress rates and enhanced mass and heat transfer. These benefits are the result of very good mixing in OFR achieved by vortex generation. OFR consists of cylindrical tube containing equally spaced orifice baffles. Fluid oscillations are superimposed on a net (laminar flow. Eddies are generated when oscillating fluid collides with baffles and passes through orifices. Generation and propagation of vortices create uniform mixing in each reactor cavity (between baffles, providing an overall flow pattern which is close to plug flow. Oscillations can be created by direct action of a piston or a diaphragm on fluid (or alternatively on baffles. This article provides an overview of oscillatory flow reactor technology, its operating principles and basic design and scale - up characteristics. Further, the article reviews the key research findings in heat and mass transfer, shear stress, residence time distribution in OFR, presenting their advantages over the conventional reactors. Finally, relevant process intensification examples from pharmaceutical, polymer and biofuels industries are presented.

  4. Neutronics and mass transport in a chemical reactor associated with controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Dang, V.D.; Steinberg, M.; Lazareth, O.W.; Powell, J.R.

    1976-05-01

    The formation of ozone from oxygen and the dissociation carbon dioxide to carbon monoxide and oxygen is studied in a gamma-neutron chemical process blanket associated with a controlled thermonuclear reactor. Materials used for reactor tube wall will affect the efficiency of the energy absorption by the reactants and consequently the yield of reaction products. Three kinds of materials, aluminum, stainless steel and fiber (Al 2 O 3 )-aluminium are investigated for the tube wall material in the study

  5. 75 FR 70351 - Termination of Environmental Review Process Cities of Chesapeake and Virginia Beach, VA

    Science.gov (United States)

    2010-11-17

    ... Virginia Beach, VA AGENCY: Federal Highway Administration (FHWA), DOT. ACTION: Termination of environmental... the Cities of Chesapeake and Virginia Beach, Virginia, is terminated. FOR FURTHER INFORMATION CONTACT...

  6. Method of operating FBR type reactors

    International Nuclear Information System (INIS)

    Arie, Kazuo.

    1984-01-01

    Purpose: To secure the controlling performance and the safety of FBR type reactors by decreasing the amount of deformation due to the difference in the heat expansion of a control rod guide tube. Method: The reactor is operated while disposing reactor core fuel assemblies of a same power at point-to-point symmetrical positions relative to the axial center for the control rod assembly. This can eliminate the temperature difference between opposing surfaces of the control rod guide tube and eliminate the difference in the thermal expansion. (Yoshino, Y.)

  7. Mechatronics education at Virginia Tech

    Science.gov (United States)

    Bay, John S.; Saunders, William R.; Reinholtz, Charles F.; Pickett, Peter; Johnston, Lee

    1998-12-01

    The advent of more complex mechatronic systems in industry has introduced new opportunities for entry-level and practicing engineers. Today, a select group of engineers are reaching out to be more knowledgeable in a wide variety of technical areas, both mechanical and electrical. A new curriculum in mechatronics developed at Virginia Tech is starting to bring students from both the mechanical and electrical engineering departments together, providing them wit an integrated perspective on electromechanical technologies and design. The course is cross-listed and team-taught by faculty from both departments. Students from different majors are grouped together throughout the course, each group containing at least one mechanical and one electrical engineering student. This gives group members the ability to learn from one another while working on labs and projects.

  8. Virginia Tech honors women in March

    OpenAIRE

    Lazenby, Jenna

    2007-01-01

    Commemorating National Women's History Month in March, the Virginia Tech community will host a variety of informative, educational, and entertaining events and programs that highlight women's diverse experiences and achievements.

  9. 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

  10. 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...

  11. Leak Detectives Saving Money, Water in Virginia

    Science.gov (United States)

    “Circuit riders” from the Virginia Rural Water Association (VRWA) are traveling to small communities across the Commonwealth using special equipment financed by EPA to locate expensive and wasteful leaks in drinking water distribution systems.

  12. The Redesign of Developmental Education in Virginia

    Science.gov (United States)

    Edgecombe, Nikki

    2016-01-01

    This chapter describes the structure and implementation of a redesign of developmental education in the Virginia Community College System, discusses preliminary descriptive findings from an evaluation of the redesign, and shares lessons for the field.

  13. 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...

  14. Team West Virginia/Rome Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Korakakis, Dimitris [West Virginia Univ., Morgantown, WV (United States)

    2017-04-10

    Overall, the team, West Virginia University (WVU) and University of Rome Tor Vergata (UTV), has a goal of building an attractive, low-cost, energy-efficient solar-powered home that represents both the West Virginian and Italian cultures.

  15. 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...

  16. DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

    Science.gov (United States)

    Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

    1962-08-14

    A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

  17. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  18. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2001-01-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

  19. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2002-04-01

    SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

  20. Reactor shutdown device

    International Nuclear Information System (INIS)

    Matsumiya, Hirohito; Endo, Hiroshi; Tsuboi, Yasushi.

    1993-01-01

    The present invention concerns a reactor shutdown device capable of suppressing change of a core insertion amount relative to temperature change during normal operation and having a great extension amount due to thermal expansion and high mechanical strength. A control rod main body is contained vertically movably in a guide tube disposed in a reactor core. An extension member extends upward from the upper end of a control rod main body and suspends the control rod main body. A shrinkable member intervenes at a midway of the extension member and is made shrinkable. A temperature sensitive member contains coolants at the inside and surrounds the shrinkable member. Thus, if the temperature of external coolants rises abruptly, the shrinkable member is extended by thermal expansion of the coolants in the temperature sensitive member. Upon usual reactor startup, the coolants in the temperature sensitive member cause no substantial thermal expansion by temperature elevation from a cold shutdown temperature to a rated power operation temperature, and the shrinkable member maintains its original state, so that the control rod main body is not inserted into the reactor core. However, upon abrupt temperature elevation, the control rod main body is inserted into the reactor core. (I.S.)

  1. Tracheostomy tube - speaking

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/patientinstructions/000465.htm Tracheostomy tube - speaking To use the sharing features on ... are even speaking devices that can help you. Tracheostomy Tubes and Speaking Air passing through vocal cords ( ...

  2. X-ray tubes

    International Nuclear Information System (INIS)

    Young, R.W.

    1979-01-01

    A form of x-ray tube is described which provides satisfactory focussing of the electron beam when the beam extends for several feet from gun to target. Such a tube can be used for computerised tomographic scanning. (UK)

  3. Neural Tube Defects

    Science.gov (United States)

    Neural tube defects are birth defects of the brain, spine, or spinal cord. They happen in the ... that she is pregnant. The two most common neural tube defects are spina bifida and anencephaly. In ...

  4. Jejunostomy feeding tube

    Science.gov (United States)

    ... may replace the tube every now and then. Cleaning the Skin Around the J-tube To clean ... this important distinction for online health information and services. Learn more about A.D.A.M.'s editorial ...

  5. Operational safety and reactor life improvements of Kyoto University Reactor

    International Nuclear Information System (INIS)

    Utsuro, M.; Fujita, Y.; Nishihara, H.

    1990-01-01

    Recent important experience in improving the operational safety and life of a reactor are described. The Kyoto University Reactor (KUR) is a 25-year-old 5 MW light water reactor provided with two thermal columns of graphite and heavy water as well as other kinds of experimental facilities. In the graphite thermal column, noticeable amounts of neutron irradiation effects had accumulated in the graphite blocks near the core. Before the possible release of the stored energy, all the graphite blocks in the column were successfully replaced with new blocks using the opportunity provided by the installation of a liquid deuterium cold neutron source in the column. At the same time, special seal mechanisms were provided for essential improvements to the problem of radioactive argon production in the column. In the heavy-water thermal column we have accomplished the successful repair of a slow leak of heavy water through a thin instrumentation tube failure. The repair work included the removal and reconstructions of the lead and graphite shielding layers and welding of the instrumentation tube under radiation fields. Several mechanical components in the reactor cooling system were also exchanged for new components with improved designs and materials. On-line data logging of almost all instrumentation signals is continuously performed with a high speed data analysis system to diagnose operational conditions of the reactor. Furthermore, through detailed investigations on critical components, operational safety during further extended reactor life will be supported by well scheduled maintenance programs

  6. Norfolk and southern eastville 10 x 20 NTMS areas Virginia and North Carolina. Data report (abbreviated)

    International Nuclear Information System (INIS)

    Cook, J.R.

    1981-06-01

    This abbreviated data report presents results of ground water and stream sediment reconnaissance in the National Topographic Map Series (NTMS) Norfolk 1 0 x 2 0 quadrangle and the southern one-half of the Eastville 1 0 x 2 0 quadrangle. Surface sediment samples were collected at 840 sites. Ground water samples were collected at 1008 sites. Neutron activation analysis (NAA) results are given for uranium and 16 other elements in sediments, and for uranium and 8 other elements in ground water. Field measurements and observations are reported for each site. Analytical data and field measurements are presented in tables and maps. Data from ground water sites include: (1) water chemistry measurements (pH, conductivity, and alkalinity); (2) physical measurements, where applicable (water temperature, well description, etc.); and (3) elemental analyses (U, Al, Br, Cl, Dy, F, Mn, Na, and V). Data from sediment sites include: (1) stream water chemistry measurements (pH, conductivity, and alkalinity); and (2) elemental analyses for sediment samples (U, Th, Hf, Al, Ce, Dy, Eu, Fe, La, Lu, Mn, Sc, Sm, Na, Ti, V, and Yb). Sample site descriptors (stream characteristics, vegetation, etc.) are also tabulated. Areal distribution maps, histograms, and cumulative frequency plots for most elements and for U/Th and U/Hf ratios are included. Uranium concentrations in the sediments that were above detection limits ranged from 0.60 to 40.2 ppM. The mean of the logarithms of the uranium concentrations was 0.61. A large area of high uranium concentrations occurs in the southwestern part of the Norfolk quadrangle. High concentrations of thorium and hafnium in the same area indicate that the uranium is associated with the resistate minerals monazite and zircon

  7. Pulsed Drift Tube Accelerator

    International Nuclear Information System (INIS)

    Faltens, A.

    2004-01-01

    The pulsed drift-tube accelerator (DTA) concept was revived by Joe Kwan and John Staples and is being considered for the HEDP/WDM application. It could be used to reach the full energy or as an intermediate accelerator between the diode and a high gradient accelerator such as multi-beam r.f. In the earliest LBNL HIF proposals and conceptual drivers it was used as an extended injector to reach energies where an induction linac with magnetic quadrupoles is the best choice. For HEDP, because of the very short pulse duration, the DTA could provide an acceleration rate of about 1MV/m. This note is divided into two parts: the first, a design based on existing experience; the second, an optimistic extrapolation. The first accelerates 16 parallel K + beams at a constant line charge density of 0.25(micro) C/m per beam to 10 MeV; the second uses a stripper and charge selector at around 4MeV followed by further acceleration to reach 40 MeV. Both benefit from more compact sources than the present 2MV injector source, although that beam is the basis of the first design and is a viable option. A pulsed drift-tube accelerator was the first major HIF experiment at LBNL. It was designed to produce a 2(micro)s rectangular 1 Ampere C s + beam at 2MeV. It ran comfortably at 1.6MeV for several years, then at lower voltages and currents for other experiments, and remnants of that experiment are in use in present experiments, still running 25 years later. The 1A current, completely equivalent to 1.8A K + , was chosen to be intermediate between the beamlets appropriate for a multi-beam accelerator, and a single beam of, say, 10A, at injection energies. The original driver scenarios using one large beam on each side of the reactor rapidly fell out of favor because of the very high transverse and longitudinal fields from the beam space charge, circa 1MV/cm and 250 kV/cm respectively, near the chamber and because of aberrations in focusing a large diameter beam down to a 1mm radius spot at a

  8. Thermal hydraulic characteristics of a double-walled tube advanced nuclear steam generator

    International Nuclear Information System (INIS)

    Cho, S.M.; Seltzer, A.H.

    1989-01-01

    In this paper the thermal hydraulic characteristics of double-walled tube steam generator designed for sodium-cooled nuclear reactors are presented. The double-walled tube construction, along with double-barrier welds for tube-to-tubesheet joints, virtually eliminates the probability of heat transfer tube failure. Considerations are given to the use of the internal core tube, helical vane swirl generator, external protector tube, and variably perforated flow baffles to improve thermal and hydraulic performance of the steam generator. These thermal hydraulic design features with a particular reference to a 432 MW PRISM steam generator are discussed

  9. Quenching behaviour of hot zircaloy tube

    International Nuclear Information System (INIS)

    Chinchole, A.S.; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.

    2015-01-01

    The quenching process plays a very important role in case of safety of nuclear reactors. During large break Loss of Coolant Accident in a nuclear reactor, the cooling water from the system is lost. Under this condition, cold water is injected from emergency core cooling system. Quenching behaviour of such heated rod bundle is really complex. It is well known that nanofluids have better heat removal capability and high heat transfer coefficient owing to enhanced thermal properties. Alumina nano-particles result in better cooling abilities compared with the traditionally used quenching media. In this paper, the authors have carried out experiments on quenching behaviour of hot zircaloy tube with demineralized water and nanofluids. It was observed that, the tube got quenched within few seconds even with the presence of decay heat and shows slightly reduced quenching time compared with DM water. (author)

  10. Reactor safeguards

    CERN Document Server

    Russell, Charles R

    1962-01-01

    Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

  11. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  12. Nasogastric and feeding tubes.

    Science.gov (United States)

    Gharib, Ahmed M; Stern, Eric J; Sherbin, Vandy L; Rohrmann, Charles A

    1996-05-01

    Preview The authors' experience in a radiology department suggested to them that there is a wide range of beliefs among practitioners regarding proper placement of nasogastric and feeding tubes. Improper positioning can cause serious problems, as they explain. Indications for different tube positions, complications of incorrect tube placement, and directions for proper positioning are discussed and illustrated.

  13. Materials for nuclear reactors

    International Nuclear Information System (INIS)

    Banerjee, S.; Kamath, H.S.

    2005-01-01

    The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO 2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO 2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

  14. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

  15. 21 CFR 868.5800 - Tracheostomy tube and tube cuff.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Tracheostomy tube and tube cuff. 868.5800 Section... (CONTINUED) MEDICAL DEVICES ANESTHESIOLOGY DEVICES Therapeutic Devices § 868.5800 Tracheostomy tube and tube cuff. (a) Identification. A tracheostomy tube and tube cuff is a device intended to be placed into a...

  16. Stress relaxation analysis and irradiation creep and swelling in pressure tubes

    International Nuclear Information System (INIS)

    Beeston, J.M.; Burr, T.K.

    1979-01-01

    An analysis is presented of slit width test information on two pressure tubes that had been irradiated in test reactors. The analysis showed that differential swelling stresses and thermal stresses undergo relaxation. The mechanism responsible for the stress relaxation at temperatures less than 700 K was irradiation creep. Irradiation creep in thermal test reactor pressure tubes is evidently greater than it would be at equivalent conditions in fast reactors. The residual stresses observed in the slit width tests varied between 30 and 257 MPa and would act to reduce the operating stresses, thus allowing for increased service life of the tubes as compared with no stress relaxation

  17. Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martens, Frederick H. [Argonne National Laboratory; Jacobson, Norman H.

    1968-09-01

    This booklet discusses research reactors - reactors designed to provide a source of neutrons and/or gamma radiation for research, or to aid in the investigation of the effects of radiation on any type of material.

  18. Nuclear reactors

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1977-01-01

    Reference is made to water cooled reactors and in particular to the cooling system of steam generating heavy water reactors (SGHWR). A two-coolant circuit is described for the latter. Full constructural details are given. (U.K.)

  19. Circumferential cracking of steam generator tubes

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    1997-04-01

    On April 28, 1995, the U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 95-03, open-quote Circumferential Cracking of Steam Generator Tubes.close-quote GL 95-03 was issued to obtain information needed to verify licensee compliance with existing regulatory requirements regarding the integrity of steam generator tubes in domestic pressurized-water reactors (PWRs). This report briefly describes the design and function of domestic steam generators and summarizes the staff's assessment of the responses to GL 95-03. The report concludes with several observations related to steam generator operating experience. This report is intended to be representative of significant operating experience pertaining to circumferential cracking of steam generator tubes from April 1995 through December 1996. Operating experience prior to April 1995 is discussed throughout the report, as necessary, for completeness

  20. [Workshop for coordinating South Carolina`s pre-college systemic initiatives in science and mathematics

    Energy Technology Data Exchange (ETDEWEB)

    1992-12-31

    On December 19, 1991, South Carolina`s Governor, established the Governor`s Mathematics and Sciences Advisory Board (MSAB) to articulate a vision and develop a statewide plan for improving science and mathematics education in South Carolina. The MSAB recognized that systemic change must occur if the achievement levels of students in South Carolina are to improve in a dramatic way. The MSAB holds two fundamental beliefs about systemic change: (1) All the elements of the science and mathematics education system must be working in harmony towards the same vision; and (2) Each element of the system must be held against high standards and progress must be assessed regularly against these standards.

  1. Reactor Neutrinos

    OpenAIRE

    Kim, Soo-Bong; Lasserre, Thierry; Wang, Yifang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  2. BOILING REACTORS

    Science.gov (United States)

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  3. Reactor vessel

    NARCIS (Netherlands)

    Makkee, M.; Kapteijn, F.; Moulijn, J.A.

    1999-01-01

    A reactor vessel (1) comprises a reactor body (2) through which channels (3) are provided whose surface comprises longitudinal inwardly directed parts (4) and is provided with a catalyst (6), as well as buffer bodies (8, 12) connected to the channels (3) on both sides of the reactor body (2) and

  4. NEI You Tube Videos: Amblyopia

    Medline Plus

    Full Text Available ... YouTube Videos >> NEI YouTube Videos: Amblyopia Listen NEI YouTube Videos YouTube Videos Home Age-Related Macular Degeneration ... Retinopathy of Prematurity Science Spanish Videos Webinars NEI YouTube Videos: Amblyopia Embedded video for NEI YouTube Videos: ...

  5. NEI You Tube Videos: Amblyopia

    Medline Plus

    Full Text Available ... YouTube Videos > NEI YouTube Videos: Amblyopia NEI YouTube Videos YouTube Videos Home Age-Related Macular Degeneration Amblyopia ... of Prematurity Science Spanish Videos Webinars NEI YouTube Videos: Amblyopia Embedded video for NEI YouTube Videos: Amblyopia ...

  6. NEI You Tube Videos: Amblyopia

    Medline Plus

    Full Text Available ... NEI YouTube Videos > NEI YouTube Videos: Amblyopia NEI YouTube Videos YouTube Videos Home Age-Related Macular Degeneration ... Retinopathy of Prematurity Science Spanish Videos Webinars NEI YouTube Videos: Amblyopia Embedded video for NEI YouTube Videos: ...

  7. Nuclear power reactors

    International Nuclear Information System (INIS)

    1982-11-01

    After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

  8. 2012 South Carolina DNR Lidar: Abbeville County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Towill Inc. collected LiDAR for over 3,300 square miles in Calhoun, Aiken, Barnwell, Edgefield, McCormick, and Abbeville counties in South Carolina. This metadata...

  9. 2008 South Carolina Lidar: Fairfield County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  10. 2008 South Carolina Lidar: Darlington County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  11. 2008 South Carolina Lidar: Marlboro County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  12. 2008 South Carolina Lidar: Marion County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  13. 2008 South Carolina Lidar: Chesterfield County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  14. 2008 South Carolina Lidar: Dillon County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  15. 2008 South Carolina Lidar: Newberry County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  16. 2008 South Carolina Lidar: Chester County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  17. 2008 South Carolina Lidar: Greenwood County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  18. 2008 South Carolina Lidar: Williamsburg County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  19. 2008 South Carolina Lidar: Union County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  20. 2008 South Carolina Lidar: Orangeburg County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  1. Boletus durhamensis sp. nov. from North Carolina

    Science.gov (United States)

    Beatriz Ortiz-Santana; Alan E. Bessette; Owen L. McConnell

    2016-01-01

    A new bolete with cinnamon-brown pores, Boletus durhamensis, is described. Collected in northern North Carolina, it is possibly mycorrhizal with Quercus spp. Morphological and molecular characters support this taxon as a new species.

  2. 2008 South Carolina Lidar: Lancaster County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The project area is composed of 16 counties in the State of South Carolina - Cherokee, Union, Laurens, Greenwood, Newberry, Chester, Fairfield, Lancaster,...

  3. 2014 Horry County, South Carolina Lidar

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set is comprised of lidar point cloud data. This project required lidar data to be acquired over Horry County, South Carolina. The total area of the Horry...

  4. 2012 South Carolina DNR Lidar: Calhoun County

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Towill Inc. collected LiDAR for over 3,300 square miles in Calhoun, Aiken, Barnwell, Edgefield, McCormick, and Abbeville counties in South Carolina. This metadata...

  5. Reactor physics and reactor computations

    International Nuclear Information System (INIS)

    Ronen, Y.; Elias, E.

    1994-01-01

    Mathematical methods and computer calculations for nuclear and thermonuclear reactor kinetics, reactor physics, neutron transport theory, core lattice parameters, waste treatment by transmutation, breeding, nuclear and thermonuclear fuels are the main interests of the conference

  6. Research reactors

    International Nuclear Information System (INIS)

    Merchie, Francois

    2015-10-01

    This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

  7. Tomographic visualization of stress corrosion cracks in tubing

    International Nuclear Information System (INIS)

    Morris, R.A.; Kruger, R.P.; Wecksung, G.W.

    1979-06-01

    A feasibility study was conducted to determine the possibility of detecting and sizing cracks in reactor cooling water tubes using tomographic techniques. Due to time and financial constraints, only one tomographic reconstruction using the best technique available was made. The results indicate that tomographic reconstructions can, in fact, detect cracks in the tubing and might possibly be capable of measuring the depth of the cracks. Limits of detectability and sensitivity have not been determined but should be investigated in any future work

  8. High-temperature transient creep properties of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Fong, R.W.L.; Chow, C.K.

    2002-06-01

    During a hypothetical large break loss-of-coolant accident (LOCA), the coolant flow would be reduced in some fuel channels and would stagnate and cause the fuel temperature to rise and overheat the pressure tube. The overheated pressure tube could balloon (creep radially) into contact with its moderator-cooled calandria tube. Upon contact, the stored thermal energy in the pressure tube is transferred to the calandria tube and into the moderator, which acts as a heat sink. For safety analyses, the modelling of fuel channel deformation behaviour during a large LOCA requires a sound knowledge of the high-temperature creep properties of Zr-2.5Nb pressure tubes. To this extent, a ballooning model to predict pressure-tube deformation was developed by Shewfelt et al., based on creep equations derived using uniaxial tensile specimens. It has been recognized, however, that there is an inherent variability in the high-temperature creep properties of CANDU pressure tubes. The variability, can be due to different tube-manufacturing practices, variations in chemical compositions, and changes in microstructure induced by irradiation during service in the reactor. It is important to quantify the variability of high-temperature creep properties so that accurate predictions on pressure-tube creep behaviour can be made. This paper summarizes recent data obtained from high-temperature uniaxial creep tests performed on specimens taken from both unirradiated (offcut) and irradiated pressure tubes, suggesting that the variability is attributed mainly to the initial differences in microstructure (grain size, shape and preferred orientation) and also from tube-to-tube variations in chemical composition, rather than due to irradiation exposure. These data will provide safety analysts with the means to quantify the uncertainties in the prediction of pressure-tube contact temperatures during a postulated large break LOCA. (author)

  9. Diagnosis of electric equipment at the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Nguyen Truong Sinh

    1999-01-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type of its kind in the world: Soviet-designed core and control system harmoniously integrated into the left-over infrastructure of the former American-made TRIGA MARK II reactor, which includes the reactor tank and shielding, graphite reflector, beam tubes and thermal column. The reactor is mainly used for radioisotope and radiopharmaceutical production, elemental analysis using neutron activation techniques, neutron beam exploitation, silicon doping, and reactor physics experimentation. For safe operation of the reactor maintenance work has been carried out for the reactor control and instrumentation, reactor cooling, ventilation, radiomonitoring, mechanical, normal electric supply systems as well as emergency electric diesel generators and the water treatment station. Technical management of the reactor includes periodical maintenance as required by technical specifications, training, re-training and control of knowledge for reactor staff. During recent years, periodic preventive maintenance (PPM) has been carried out for the electric machines of the technological systems. (author)

  10. Nuclear reactor plant

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1977-01-01

    The invention is concerned with a quick-closing valve on the main-steam pipe of a nuclear reactor plant. The quick-closing valve serves as isolating valve and as safety valve permitting depressurization in case of an accident. For normal operation a tube-shaped gate valve is provided as valve disc, enclosing an auxiliary valve disc to be used in case of accidents and which is opened at increased pressure to provide a smaller flow cross-section. The design features are described in detail. (RW) [de

  11. Atomic hydrogen reactor

    International Nuclear Information System (INIS)

    Massip de Turville, C.M.D.

    1982-01-01

    Methods are discussed of generating heat in an atomic hydrogen reactor which involve; the production of atomic hydrogen by an electrical discharge, the capture of nascent neutrons from atomic hydrogen in a number of surrounding steel alloy tubes having a high manganese content to produce 56 Mn, the irradiation of atomic hydrogen by the high energy antineutrinos from the beta decay of 56 Mn to yield nascent neutrons, and the removal of the heat generated by the capture of nascent neutrons by 55 Mn and the beta decay of 56 Mn. (U.K.)

  12. Safety issues at the defense production reactors

    International Nuclear Information System (INIS)

    1987-01-01

    The United States produces plutonium and tritium for use in nuclear weapons at the defense production reactors endash the N Reactor in Washington and the Savannah River reactors in South Carolina. This report reaches general conclusions about the management of those reactors and highlights a number of safety and technical issues that should be resolved. The report provides an assessment of the safety management, safety review, and safety methodology employed by the Department of Energy and the private contractors who operate the reactors for the federal government. The report is necessarily based on a limited review of the defense production reactors. It does not address whether any of the reactors are ''safe,'' because such an analysis would involve a determination of acceptable risk endash a matter of obvious importance, but one that was beyond the purview of the committee. It also does not address whether the safety of the production reactors is comparable to that of commercial nuclear power stations, because even this narrower question extended beyond the charge to the committee and would have involved detailed analyses that the committee could not undertake

  13. Education activities of the Eastern Carolina Section

    International Nuclear Information System (INIS)

    Eckenrode, M.; Hudson, O.N.

    1991-01-01

    The Eastern Carolinas Section (ECS) Education Committee has successfully tapped into all grade levels of eastern North Carolina's public schools. The keys to access are building a section organization geared toward education and maintaining a wide variety of products from which teachers can choose. The education committee conservatively estimates that in 1990 it relayed information on nuclear-related issues to over a thousand students

  14. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Nagatomi, Shozo.

    1976-01-01

    Object: To provide a jet and missile protective wall of a configuration being inflated toward the center of a reactor container on the inside of a body of the reactor container disposed within a biological shield wall to thereby increase safety of the reactor container. Structure: A jet and missile protective wall comprised of curved surfaces internally formed with a plurality of arch inflations filled with concrete between inner and outer iron plates and shape steel beam is provided between a reactor container surrounded by a biological shield wall and a thermal shield wall surrounding the reactor pressure vessel, and an adiabatic heat insulating material is filled in space therebetween. (Yoshino, Y.)

  15. Virginia Beach Tsunami Forecast Grids for MOST Model

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Virginia Beach, Virginia Forecast Model Grids provides bathymetric data strictly for tsunami inundation modeling with the Method of Splitting Tsunami (MOST)...

  16. 77 FR 69489 - Virginia; Emergency and Related Determinations

    Science.gov (United States)

    2012-11-19

    ... determined that the emergency conditions in the Commonwealth of Virginia resulting from Hurricane Sandy... Commonwealth of Virginia have been designated as adversely affected by this declared emergency: Emergency... 69490

  17. LIGHT WATER MODERATED NEUTRONIC REACTOR

    Science.gov (United States)

    Christy, R.F.; Weinberg, A.M.

    1957-09-17

    A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

  18. Damping of reactor internals

    International Nuclear Information System (INIS)

    Singleton, N.R.; Bohm, G.J.

    1977-01-01

    This paper presents and discusses the results of a study of internals damping using data obtained from wave analysis of PWR plant flow tests, and shaker tests. The damping values were obtained from vibration data taken during the pre-operational testing of several reactor plants and some in air shaker tests. Parameters available in the data include the presence of the core, the presence and position of the control rod drive line, reactor coolant temperature, and combination of reactor coolant pumps in operation. Modal damping values for the structures reported were obtained from the decay of autocorrellograms or from the modal response half-power bandwidths of frequency spectra. For the lower frequency core barrel-reactor vessel beam modes 2% to 5% damping values were found for minimum damping values. Significantly higher values are found in the data when, for example, intermittent contact occurs at the core barrel level supports. Core barrel and thermal shield shell modes having natural frequencies in the frequency range of interest for seismic response calculations exhibit damping values generally on the order of 1% to 2%. Higher frequency, very low amplitude, shell modes of these structures can have damping values of less than 1%. Damping values for guide tubes were found to have minimum values of 2% to 5% depending on their core location. The cross flow velocity and thus the floor turbulence excited amplitude is higher for guide tubes in core locations near the outlet nozzles. Information on the damping of upper support columns which are similarly excited is also given. Damping values reported are suitable for normal operation design conditions, i.e., for oscillatory behavior and relatively small amplitudes. The extrapolation of the data to obtain realistic values for large seismic events and for loss of coolant accidents is also discussed

  19. Stitching the western Piedmont of Virginia: Early Paleozoic tectonic history of the Ellisville Pluton and the Potomac and Chopawamsic Terranes

    Science.gov (United States)

    Hughes, K. S.; Hibbard, J. P.; Sauer, R.T.; Burton, William C.

    2014-01-01

    The theme of the 2014 Virginia Geological Field Conference is the tectonic development, economic geology, and seismicity of the western Piedmont of Louisa County, Virginia. It is timely for the conference to turn its attention here, for during the past decade these aspects of western Piedmont geology have garnered the renewed attention of researchers. In terms of regional tectonics, it has been hypothesized that the major structure in the region, the Chopawamsic fault system, represents the most significant boundary in the Appalachian orogen, the main Iapetan suture (Hibbard et al., 2014). Economically, recent elevated market values of metals— particularly that of gold—has spurred reconsideration of the economic geology of the western Piedmont. Finally, the August 23, 2011, M5.8 earthquake, with its epicenter in our field area, startled the North American east coast and has revived awareness of the seismic potential of the region. This renewed interest in the geology of the western Piedmont of north-central Virginia has led to new detailed bedrock mapping, detailed surficial mapping, high-resolution UPb TIMS zircon geochronology, U-Pb LA-ICPMS detrital zircon geochronology, radiogenic isotope geochemistry, major/minor/REE geochemistry, and geophysical studies (e.g. Bailey et al., 2005, 2008; Bailey and Owens, 2012: Berti et al., 2012; Burton et al., 2014; Burton, in progress; Harrison, 2012; Horton et al., 2010, in press; Hughes, 2010, 2014; Hughes et al., 2013a, 2013b, 2014, in press a, in press b; Malenda, in progress; Owens et al., 2013; Spears and Gilmer 2012; Spears et al. 2013, Terblanche, 2013; Terblanche and Nance, 2012). A host of institutions have taken part in the research, including North Carolina State University, the Virginia Department of Mines, Minerals, and Energy, the U.S. Geological Survey, Virginia Tech, Lehigh University, and the College of William and Mary. Many of these investigations remain active. The majority of the data presented

  20. Calculation on a tube bundle: change in bundle stiffness; Calcul sur un faisceau de tubes: changement de raideur au sein du faisceau

    Energy Technology Data Exchange (ETDEWEB)

    Goasdoue, J

    1998-06-01

    Tube bundles immersed in a fluid are frequently used in nuclear equipment like reactor core. In this work, tubes have not the same stiffness. The aim of this study is then to analyze the influence of stiffness on the structure behaviour. The CASTEM 2000 code has been used. (O.M.)

  1. Numerical analysis of an experimental data base for tubes pulled in flexion; Analyse numerique d'une base de donnees experimentales de tubes sollicites en flexion

    Energy Technology Data Exchange (ETDEWEB)

    Langlois, R

    1998-07-01

    The aim of this study is the simulation and the interpretation of experimental results about maximal loading that tubes are able to carry. The tubes are products from primary circuit of german power reactors light water moderated boiling and not boiling cooled. The crack propagation is evaluate under loading. (A.L.B.)

  2. Reactor instrumentation and control

    International Nuclear Information System (INIS)

    Wach, D.; Beraha, D.

    1980-01-01

    The methods for measuring radiation are shortly reviewed. The instrumentation for neutron flux measurement is classified into out-of-core and in-core instrumentation. The out-of-core instrumentation monitors the operational range from the subcritical reactor to full power. This large range is covered by several measurement channels which derive their signals from counter tubes and ionization chambers. The in-core instrumentation provides more detailed information on the power distribution in the core. The self-powered neutron detectors and the aeroball system in PWR reactors are discussed. Temperature and pressure measurement devices are briefly discussed. The different methods for leak detection are described. In concluding the plant instrumentation part some new monitoring systems and analysis methods are presented: early failure detection methods by noise analysis, acoustic monitoring and vibration monitoring. The presentation of the control starts from an qualitative assessment of the reactor dynamics. The chosen control strategy leads to the definition of the part-load diagram, which provides the set-points for the different control systems. The tasks and the functions of these control systems are described. In additiion to the control, a number of limiting systems is employed to keep the reactor in a safe operating region. Finally, an outlook is given on future developments in control, concerning mainly the increased application of process computers. (orig./RW)

  3. Nuclear Reactor Sharing Program

    International Nuclear Information System (INIS)

    1994-01-01

    The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making reactor parameter measurements. For neutron activation analysis and analyses of natural environmental radioactivity, the NRL maintains the gamma ray spectroscopy system (GRSS). It is comprised of two PC-based 8192-channel multichannel analyzers (MCAs) with all the required software for quantitative analysis. A 3 double-prime x 3 double-prime NaI(Tl), a 14 percent Ge(Li), and a High Purity Germanium detector are currently available for use with the spectroscopy system

  4. Physics and safety of advanced research reactors

    International Nuclear Information System (INIS)

    Boening, K.; Hardt, P. von der

    1987-01-01

    Advanced research reactor concepts are presently being developed in order to meet the neutron-based research needs of the nineties. Among these research reactors, which are characterized by an average power density of 1-10 MW per liter, highest priority is now generally given to the 'beam tube reactors'. These provide very high values of the thermal neutron flux (10 14 -10 16 cm -2 s -1 ) in a large volume outside of the reactor core, which can be used for sample irradiations and, in particular, for neutron scattering experiments. The paper first discusses the 'inverse flux trap concept' and the main physical aspects of the design and optimization of beam tube reactors. After that two examples of advanced research reactor projects are described which may be considered as two opposite extremes with respect to the physical optimization principle just mentioned. The present situation concerning cross section libraries and neutronic computer codes is more or less satisfactory. The safety analyses of advanced research reactors can largely be updated from those of current new designs, partially taking advantage of the immense volume of work done for power reactors. The paper indicates a few areas where generic problems for advanced research reactor safety are to be solved. (orig.)

  5. Research reactors

    International Nuclear Information System (INIS)

    Kowarski, L.

    1955-01-01

    It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

  6. Lunar Lava Tube Sensing

    Science.gov (United States)

    York, Cheryl Lynn; Walden, Bryce; Billings, Thomas L.; Reeder, P. Douglas

    1992-01-01

    Large (greater than 300 m diameter) lava tube caverns appear to exist on the Moon and could provide substantial safety and cost benefits for lunar bases. Over 40 m of basalt and regolith constitute the lava tube roof and would protect both construction and operations. Constant temperatures of -20 C reduce thermal stress on structures and machines. Base designs need not incorporate heavy shielding, so lightweight materials can be used and construction can be expedited. Identification and characterization of lava tube caverns can be incorporated into current precursor lunar mission plans. Some searches can even be done from Earth. Specific recommendations for lunar lava tube search and exploration are (1) an Earth-based radar interferometer, (2) an Earth-penetrating radar (EPR) orbiter, (3) kinetic penetrators for lunar lava tube confirmation, (4) a 'Moon Bat' hovering rocket vehicle, and (5) the use of other proposed landers and orbiters to help find lunar lava tubes.

  7. An evaluation of reclamation tree planting in southwest Virginia

    Science.gov (United States)

    Danny R. Brown; Donald L. Branson

    1980-01-01

    Surface mining began in Southwest Virginia in the mid-1940's, however, few or no records were kept before 1950 and it was not until 1966 that the first Virginia reclamation law was passed. Two years later in 1968, the Division of Mined Land Reclamation was created and tree planting began on a uniform basis in Virginia surface mines.

  8. A Survey of Perceptions of the Virginia Tech Tragedy

    Science.gov (United States)

    Fallahi, Carolyn R.; Austad, Carol Shaw; Fallon, Marianne; Leishman, Lisa

    2009-01-01

    The recent shootings at the Virginia Polytechnic Institute (Virginia Tech) shocked the nation and brought violence on college campuses to the forefront of the nation's attention. We surveyed college students and faculty/staff three weeks after the incident about their perceptions of the Virginia Tech shooting, subsequent media exposure, and school…

  9. An Examination of Regional Hardwood Roundwood Markets in West Virginia

    Science.gov (United States)

    William Luppold; Delton Alderman; Delton Alderman

    2005-01-01

    West Virginia?s hardwood resource is large and diverse ranging from oak-hickory forests in the southern and western portions of the state to northern hardwood stands in the northeastern region. West Virginia also has a diverse group of primary hardwood- processing industries, including hardwood grade mills, industrial hardwood sawmills, engineered wood-product...

  10. West Virginia Interpretive Guide Training: A Collaborative Effort

    Science.gov (United States)

    Balcarczyk, Kelly; McKenney, Kathryn; Smaldone, Dave; Arborgast, Doug

    2013-01-01

    West Virginia University's Extension Service partnered with the Recreation, Parks, and Tourism Resources Program to improve guide performance in West Virginia's tourism industry. The result of this partnership is a West Virginia Interpretive Guide Training program aimed at providing low-cost, widely available training to guides throughout the…

  11. William Knocke receives 2008 Virginia Outstanding Civil Engineer Award

    OpenAIRE

    Daniilidi, Christina

    2008-01-01

    William R. Knocke, W.C. English Professor and head of the Charles E. Via, Jr. Department of Civil and Environmental Engineering at Virginia Tech, was awarded the 2008 Virginia Outstanding Civil Engineer Award at the Virginia Section of the American Society of Civil Engineers' (ASCE) banquet, held recently in Williamsburg, Va.

  12. Refueling system for a nuclear reactor

    International Nuclear Information System (INIS)

    Koschkin, J.N.; Ordynskij, G.V.; Schchijan, C.G.; Schapkin, A.F.; Fadeev, A.I.; Laptev, F.V.; Batjukov, V.I.; Korolkov, K.I.; Borodin, I.V.; Tschernomordik, E.N.

    1979-01-01

    With the refueling system fuel elements are transferred from the intermediate distributing chamber within the fast breeder reactor vessel to the storage tanks for new and irradiated fuel elements outside of the reactor vessel and vice versa. It consists of a hermetic chamber, filled with inert gas, within which the refueling machine, having got a vertical swing pipe, is placed. On the swing pipe there is mounted by means of a bracket a hanging support tube for a tube manipulator that can be moved over the openings to the fuel elements. At the end of the tube manipulator there is a gripping device whose drive mechanism is arranged within the support tube. The swing pipe is moved by means of a drive mechanism outside of the chamber. (DG) [de

  13. Manual tube welding torch

    International Nuclear Information System (INIS)

    Kiefer, J.H.; Smith, D.J.

    1981-01-01

    In a welding torch which fits over a tube intermediate the ends thereof for welding the juncture between the tube and a boss on the back side of a tube plate, a split housing encloses a tungsten electrode, a filler wire duct and a fiber optic bundle arranged to observe the welding process. A shielding gas duct is provided in the housing. A screw is provided for setting electrode/work distance. Difficult remote tube welding operations can be performed with the apparatus. (author)

  14. Wound tube heat exchanger

    Science.gov (United States)

    Ecker, Amir L.

    1983-01-01

    What is disclosed is a wound tube heat exchanger in which a plurality of tubes having flattened areas are held contiguous adjacent flattened areas of tubes by a plurality of windings to give a double walled heat exchanger. The plurality of windings serve as a plurality of effective force vectors holding the conduits contiguous heat conducting walls of another conduit and result in highly efficient heat transfer. The resulting heat exchange bundle is economical and can be coiled into the desired shape. Also disclosed are specific embodiments such as the one in which the tubes are expanded against their windings after being coiled to insure highly efficient heat transfer.

  15. Sapphire tube pressure vessel

    Science.gov (United States)

    Outwater, John O.

    2000-01-01

    A pressure vessel is provided for observing corrosive fluids at high temperatures and pressures. A transparent Teflon bag contains the corrosive fluid and provides an inert barrier. The Teflon bag is placed within a sapphire tube, which forms a pressure boundary. The tube is received within a pipe including a viewing window. The combination of the Teflon bag, sapphire tube and pipe provides a strong and inert pressure vessel. In an alternative embodiment, tie rods connect together compression fittings at opposite ends of the sapphire tube.

  16. Fuel nozzle tube retention

    Energy Technology Data Exchange (ETDEWEB)

    Cihlar, David William; Melton, Patrick Benedict

    2017-02-28

    A system for retaining a fuel nozzle premix tube includes a retention plate and a premix tube which extends downstream from an outlet of a premix passage defined along an aft side of a fuel plenum body. The premix tube includes an inlet end and a spring support feature which is disposed proximate to the inlet end. The premix tube extends through the retention plate. The spring retention feature is disposed between an aft side of the fuel plenum and the retention plate. The system further includes a spring which extends between the spring retention feature and the retention plate.

  17. Tig welding of stainless steel AISI 316 tubes for fuel rods

    International Nuclear Information System (INIS)

    Siqueira Queiroz Bittencourt, M. de.

    1985-01-01

    Sealing of austenitic stainless steel AISI 316 tubes (20% cold worked). By welding end-caps material was studied, aiming their utilization as fuel rods for nuclear reactors. It was used the autogenous TIG welding process. (author)

  18. Optimal Speed Limits for School Buses on Virginia Highways: A Report to Virginia's Superintendent of Public Instruction.

    Science.gov (United States)

    Jernigan, Jack D.; Lynn, Cheryl W.

    A study to assess whether the school bus speed limit should be changed in Virginia is described in this report. The relationship between the safety characteristics of Virginia's three-tiered speed limit system and school bus operation is examined to determine the optimal level of safety for school bus travel. Virginia allows the following three…

  19. Corium Configuration and Penetration Tube Failure for Fukushima Daiichi Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    An, Sang Mo; Lee, Jae Bong; Kim, Hwan Yeol; Song, Jin Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    For the LWRs (light water reactors), the penetration tubes at the reactor vessel lower head are regarded as the most vulnerable structures along with a global vessel failure during a severe accident because they can be seriously damaged by a corium melt or debris relocated into the lower plenum of the vessel. The research on the penetration tube failure is of higher importance in the BWRs, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. There are more than one hundred of penetration tubes in the Fukushima Daiichi NPPs (nuclear power plants), such as ICM-GTs (in-core monitoring guide tubes), CRGTs (control rod guide tubes) and drain tubes. The ICM-GTs include SRMs (source range monitors), IRMs (intermediate range monitors), LPRMs (local power range monitors) and TIPs (traversing in-core probes), which are much thinner than other tubes. The experimental researches to investigate the corium configuration and the penetration tube failure for the Fukushima Daiichi NPPs were introduced and some meaningful results were summarized. It was shown that the corium ingot was separated into two layers, of which the upper layer was metal-rich while the lower one was oxide-rich. It seemed that B{sub 4}C would contribute to reducing the density of the metallic melt. The two-layered configuration will provide useful information to understand the core melt progression and post-recovery actions for the Fukushima Daiichi NPPs. In addition, we performed a large scale penetration tube failure experiment for the SRM/IRM guide tube, and showed high possibilities of large amount of corium discharge out of the reactor vessel lower head, which followed by the tube melting in a very short time. We are planning to perform the penetration tube failure experiments for another dry tube of ICM-GT (LPRM guide tube), and later for the wet tube (CRGT)

  20. Creep behavior of materials for high-temperature reactor application

    International Nuclear Information System (INIS)

    Schneider, K.; Hartnagel, W.; Iischner, B.; Schepp, P.

    1984-01-01

    Materials for high-temperature gas-cooled reactor (HTGR) application are selected according to their creep behavior. For two alloys--Incoloy-800 used for the live steam tubing of the thorium high-temperature reactor and Inconel-617 evaluated for tubings in advanced HTGRs--creep curves are measured and described by equations. A microstructural interpretation is given. An essential result is that nonstable microstructures determine the creep behavior

  1. Characteristics of the JRR-3M neutron guide tubes

    International Nuclear Information System (INIS)

    Suzuki, Masatoshi; Ichikawa, Hiroki; Kawabata, Yuji.

    1993-01-01

    Large scale neutron guide tubes have been installed in the upgraded JRR-3 (Japan Research Reactor No.3, JRR-3M). The total length of the guide tubes is 232m. The neutron fluxes and spectra were measured at the end of the neutron guide tubes. The neutron fluxes of thermal neutron guide tubes with characteristic wavelength of 2A are 1.2 x 10 8 n/cm 2 · s. The neutron fluxes of cold guide tubes are 1.4 x 10 8 n/cm 2 · s with characteristic wavelength of 4A and 2.0 x 10 8 n/cm 2 · s with 6A when the cold neutron source is operated. The neutron spectra measured by time-of-flight method agree well with their designed ones. (author)

  2. Hybrid reactors

    International Nuclear Information System (INIS)

    Moir, R.W.

    1980-01-01

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of 233 U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m -2 , and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid

  3. Profiling a reactor component using ultrasonics

    Energy Technology Data Exchange (ETDEWEB)

    Pathak, L.; Seshadri, V.R.; Kumaravadivelu, C.; Sreenivasan, G.; Raghunathan, V.S. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-04-01

    Nuclear reactors have many components within the reactor vessel. During the life of a reactor it is possible for these components to be displaced or deformed because of the thermal cycles to which they are subject. Also, these components in situ therefore becomes important for the upkeep of the reactor. However, high radiation levels make it difficult to monitor using optical methods. This paper describes an ultrasonic method which was successfully employed in profiling a deformed guide tube of a reactor. The method uses the well-known ultrasonic ranging technique. However, the specialty of the method is the use of air transducers at 40 kHz to overcome the inherent divergence problems and the difficulties associated with high temperatures inherent in a sodium cooled reactor.

  4. Mixing Waters and Moving Ships off the North Carolina Coast

    Science.gov (United States)

    2000-01-01

    The estuarine and marine environments of the United States' eastern seaboard provide the setting for a variety of natural and human activities associated with the flow of water. This set of Multi-angle Imaging SpectroRadiometer images from October 11, 2000 (Terra orbit 4344) captures the intricate system of barrier islands, wetlands, and estuaries comprising the coastal environments of North Carolina and southern Virginia. On the right-hand side of the images, a thin line of land provides a tenuous separation between the Albemarle and Pamlico Sounds and the Atlantic Ocean. The wetland communities of this area are vital to productive fisheries and water quality.The top image covers an area of about 350 kilometers x 260 kilometers and is a true-color view from MISR's 46-degree backward-looking camera. Looking away from the Sun suppresses glint from the reflective water surface and enables mapping the color of suspended sediments and plant life near the coast. Out in the open sea, the dark blue waters indicate the Gulf Stream. As it flows toward the northeast, this ocean current presses close to Cape Hatteras (the pointed cape in the lower portion of the images), and brings warm, nutrient-poor waters northward from equatorial latitudes. North Carolina's Outer Banks are often subjected to powerful currents and storms which cause erosion along the east-facing shorelines. In an effort to save the historic Cape Hatteras lighthouse from the encroaching sea, it was jacked out of the ground and moved about 350 meters in 1999.The bottom image was created with red band data from the 46-degree backward, 70-degree forward, and 26-degree forward cameras displayed as red, green, and blue, respectively. The color variations in this multi-angle composite indicate different angular (rather than spectral) signatures. Here, the increased reflection of land vegetation at the angle viewing away from the Sun causes a reddish tint. Water, on the other hand, appears predominantly in shades

  5. Wood power in North Carolina

    Energy Technology Data Exchange (ETDEWEB)

    Cleland, J.G.; Guessous, L. [Research Triangle Institute, Research Triangle Park, NC (United States)

    1993-12-31

    North Carolina (NC) is one of the most forested states, and supports a major wood products industry. The NC Department of Natural Resources sponsored a study by Research Triangle Institute to examine new, productive uses of the State`s wood resources, especially electric power generation by co-firing with coal. This paper summarizes our research of the main factors influencing wood power generation opportunities, i.e., (1) electricity demand; (2) initiative and experience of developers; (3) available fuel resources; (4) incentives for alternate fuels; and (5) power plant technology and economics. The results cover NC forests, short rotation woody crops, existing wood energy facilities, electrical power requirements, and environmental regulations/incentives. Quantitative assessments are based on the interests of government agencies, utilities, electric cooperatives, developers and independent power producers, forest products industries, and the general public. Several specific, new opportunities for wood-to-electricity in the State are identified and described. Comparisons are made with nationwide resources and wood energy operations. Preferred approaches in NC are co-generation in existing or modified boilers and in dedicated wood power plants in forest industry regions. Co-firing is mainly an option for supplementing unreliable primary fuel supplies to existing boilers.

  6. Reactor fuel rod

    International Nuclear Information System (INIS)

    Inui, Mitsuhiro; Mori, Kazuma.

    1990-01-01

    In a high burnup degree reactor core, a problem of fuel can corrosion caused by coolants occurs due to long stay in a reactor. Then, the use of fuel cladding tubes with improved corrosion resistance is now undertaken and use of corrosion resistant alloys is attempted. However, since the conventional TIG welding melts the entire portion, the welded portion does not remain only in the corrosive resistant alloy but it forms new alloys of the corrosion resistant alloy and zircaloy as the matrix material or inter-metallic compounds, which degrades the corrosion resistance. In the present invention, a cladding tube comprising a dual layer structure using a corrosion resistant alloy only for a required thickness and an end plug made of the same material as the corrosion resistant alloy are welded at the junction portion by using resistance welding. Then, they are joined under welding by the heat generated to the junction surfaces between both of them, to provide corrosion resistant alloys substantially at the outside of the welded portion as well. Accordingly, the corrosion resistance is not degradated. (T.M.)

  7. Post-hurricane Joaquin Coastal Oblique Aerial Photographs Collected from the South Carolina/North Carolina Border to Montauk Point, New York, October 7–9, 2015

    Science.gov (United States)

    Morgan, Karen L.M.

    2016-06-27

    The U.S. Geological Survey (USGS), as part of the National Assessment of Coastal Change Hazards project, conducts baseline and storm-response photography missions to document and understand the changes in vulnerability of the Nation's coasts to extreme storms (Morgan, 2009). On October 7–9, 2015, the USGS conducted an oblique aerial photographic survey of the coast from the South Carolina/North Carolina border to Montauk Point, New York (fig. 1), aboard a Cessna 182 (aircraft) at an altitude of 500 feet (ft) and approximately 1,200 ft offshore fig. 2. This mission was conducted to collect post-Hurricane Joaquin data for assessing incremental changes in the beach and nearshore area since the last surveys, mission flown in September 2014 (Virginia to New York: Morgan, 2015), November 2012 (northern North Carolina: Morgan and others, 2014) and May 2008 (southern North Carolina: unpublished report), and the data can be used to assess of future coastal change.The photographs in this report are Joint Photographic Experts Group (JPEG) images. ExifTool was used to add the following to the header of each photo: time of collection, Global Positioning System (GPS) latitude, GPS longitude, keywords, credit, artist (photographer), caption, copyright, and contact information. The photograph locations are an estimate of the position of the aircraft at the time the photograph was taken and do not indicate the location of any feature in the images (see the Navigation Data page). These photographs document the state of the barrier islands and other coastal features at the time of the survey. Pages containing thumbnail images of the photographs, referred to as contact sheets, were created in 5-minute segments of flight time. These segments can be found on the Photos and Maps page. Photographs can be opened directly with any JPEG-compatible image viewer by clicking on a thumbnail on the contact sheet.In addition to the photographs, a Google Earth Keyhole Markup Language (KML) file

  8. Heterogeneous reactors

    International Nuclear Information System (INIS)

    Moura Neto, C. de; Nair, R.P.K.

    1979-08-01

    The microscopic study of a cell is meant for the determination of the infinite multiplication factor of the cell, which is given by the four factor formula: K(infinite) = n(epsilon)pf. The analysis of an homogeneous reactor is similar to that of an heterogeneous reactor, but each factor of the four factor formula can not be calculated by the formulas developed in the case of an homogeneous reactor. A great number of methods was developed for the calculation of heterogeneous reactors and some of them are discussed. (Author) [pt

  9. Heat exchanger for reactor core and the like

    Science.gov (United States)

    Kaufman, Jay S.; Kissinger, John A.

    1986-01-01

    A compact bayonet tube type heat exchanger which finds particular application as an auxiliary heat exchanger for transfer of heat from a reactor gas coolant to a secondary fluid medium. The heat exchanger is supported within a vertical cavity in a reactor vessel intersected by a reactor coolant passage at its upper end and having a reactor coolant return duct spaced below the inlet passage. The heat exchanger includes a plurality of relatively short length bayonet type heat exchange tube assemblies adapted to pass a secondary fluid medium therethrough and supported by primary and secondary tube sheets which are releasibly supported in a manner to facilitate removal and inspection of the bayonet tube assemblies from an access area below the heat exchanger. Inner and outer shrouds extend circumferentially of the tube assemblies and cause the reactor coolant to flow downwardly internally of the shrouds over the tube bundle and exit through the lower end of the inner shroud for passage to the return duct in the reactor vessel.

  10. Chest Tube Thoracostomy

    Science.gov (United States)

    ... in the space around the lungs (called a pleural effusion) . A chest tube may also be needed when a patient has ... or chest CT are also done to evaluate pleural fluid. If the X-ray shows a need for a chest tube to drain fluid or air, the procedure is ...

  11. Ventricular tachycardia following tube thoracotomy.

    Science.gov (United States)

    Hibbert, Benjamin; Lim, Toon Wei; Hibbert, Rebecca; Green, Martin; Gollob, Michael H; Davis, Darryl R

    2010-10-01

    Arrhythmias provoked by tube thoracotomy are a rare complication. We report a ventricular tachycardia after chest tube insertion for a device-related pneumothorax. Sinus rhythm was restored only by removal of the chest tube and insertion of a pliable pleural drain. Identification of the chest tube as an arrhythmic trigger following tube thoracotomy is essential in definitive management of refractory arrhythmias.

  12. North Carolina Statewide Lidar DEM 2015 Phase 3

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Geographic Extent: North Carolina Area of Interest, covering approximately 7,197 square miles. Dataset Description: The North Carolina LiDAR project called for the...

  13. North Carolina Statewide Lidar DEM 2014 Phase 1

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Geographic Extent: North Carolina Area of Interest for Sandy, covering approximately 9,396 square miles. Dataset Description: The North Carolina - Sandy LiDAR...

  14. DIGITAL FLOOD INSURANCE RATE MAP DATABASE, LENOIR COUNTY, NORTH CAROLINA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — This Flood Insurance Study was produced through a cooperative partnership between the State of North Carolina and FEMA. The North Carolina Floodplain Mapping...

  15. 2015 NCFMP Lidar: Statewide North Carolina (Phase 3)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Geographic Extent: North Carolina Area of Interest, covering approximately 7,197 square miles. Dataset Description: The North Carolina LiDAR project called for the...

  16. DIGITAL FLOOD INSURANCE RATE MAP DATABASE, GREENE COUNTY, NORTH CAROLINA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — This Flood Insurance Study was produced through a cooperative partnership between the State of North Carolina and FEMA. The North Carolina Floodplain Mapping...

  17. 2014 NCFMP Lidar: Statewide North Carolina (Phase 1)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Geographic Extent: North Carolina Area of Interest for Sandy, covering approximately 9,396 square miles. Dataset Description: The North Carolina - Sandy LiDAR...

  18. DIGITAL FLOOD INSURANCE RATE MAP DATABASE, HALIFAX COUNTY, NORTH CAROLINA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — This Flood Insurance Study was produced through a cooperative partnership between the State of North Carolina and FEMA. The North Carolina Floodplain Mapping...

  19. DIGITAL FLOOD INSURANCE RATE MAP DATABASE, Scotland County, North Carolina

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — This Flood Insurance Study was produced through a cooperative partnership between the State of North Carolina and FEMA. The North Carolina Floodplain Mapping...

  20. DIGITAL FLOOD INSURANCE RATE MAP DATABASE, WILSON COUNTY, NORTH CAROLINA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — This Flood Insurance Study was produced through a cooperative partnership between the State of North Carolina and FEMA. The North Carolina Floodplain Mapping...

  1. DIGITAL FLOOD INSURANCE RATE MAP DATABASE, Franklin County, NORTH CAROLINA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — This Flood Insurance Study was produced through a cooperative partnership between the State of North Carolina and FEMA. The North Carolina Floodplain Mapping...

  2. DIGITAL FLOOD INSURANCE RATE MAP DATABASE, EDGECOMBE COUNTY, NORTH CAROLINA

    Data.gov (United States)

    Federal Emergency Management Agency, Department of Homeland Security — This Flood Insurance Study was produced through a cooperative partnership between the State of North Carolina and FEMA. The North Carolina Floodplain Mapping...

  3. U.S. Geological Survey Virginia and West Virginia Water Science Center

    Science.gov (United States)

    Jastram, John D.

    2017-08-22

    The U.S. Geological Survey (USGS) serves the Nation by providing reliable scientific information to describe and understand the Earth; minimize loss of life and property from natural disasters; manage water, biological, energy, and mineral resources; and enhance and protect our quality of life. In support of this mission, the USGS Virginia and West Virginia Water Science Center works in cooperation with many entities to provide reliable, impartial scientific information to resource managers, planners, and the public.

  4. Novel oscillatory flow reactors for biotechnological applications

    OpenAIRE

    Reis, N.

    2006-01-01

    Tese de Doutoramento em Engenharia Química e Biológica This thesis explores the biotechnological applications of two novel scale-down oscillatory flow reactors (OFRs). A micro-bioreactor (working mostly in batch) and a continuous meso-reactor systems were developed based on a 4.4 mm internal diameter tube with smooth periodic constrictions (SPC), both operating under oscillatory flow mixing (OFM). The first part is dedicated to the flow characterisation in the novel SPC geom...

  5. Molybdenum Tube Characterization report

    Energy Technology Data Exchange (ETDEWEB)

    Beaux II, Miles Frank [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Usov, Igor Olegovich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-07

    Chemical vapor deposition (CVD) techniques have been utilized to produce free-standing molybdenum tubes with the end goal of nuclear fuel clad applications. In order to produce tubes with properties desirable for this application, deposition rates were lowered requiring long deposition durations on the order of 50 hours. Standard CVD methods as well as fluidized-bed CVD (FBCVD) methods were applied towards these objectives. Characterization of the tubes produced in this manner revealed material suitable for fuel clad applications, but lacking necessary uniformity across the length of the tubes. The production of freestanding Mo tubes that possess the desired properties across their entire length represents an engineering challenge that can be overcome in a next iteration of the deposition system.

  6. Categorising YouTube

    DEFF Research Database (Denmark)

    Simonsen, Thomas Mosebo

    2011-01-01

    This article provides a genre analytical approach to creating a typology of the User Generated Content (UGC) of YouTube. The article investigates the construction of navigation processes on the YouTube website. It suggests a pragmatic genre approach that is expanded through a focus on YouTube’s...... technological affordances. Through an analysis of the different pragmatic contexts of YouTube, it is argued that a taxonomic understanding of YouTube must be analysed in regards to the vacillation of a user-driven bottom-up folksonomy and a hierarchical browsing system that emphasises a culture of competition...... and which favours the already popular content of YouTube. With this taxonomic approach, the UGC videos are registered and analysed in terms of empirically based observations. The article identifies various UGC categories and their principal characteristics. Furthermore, general tendencies of the UGC within...

  7. Some critical corrosion issues and mitigation strategies affecting light water reactors

    International Nuclear Information System (INIS)

    Jones, R.L.

    1996-01-01

    Recent corrosion experience in US light water reactor nuclear power plants is reviewed with emphasis on mitigation strategies to control the cost of corrosion to light water reactor operators. The most costly issues have been stress corrosion cracking of stainless steel coolant piping in boiling water reactors and corrosion damage to steam generator tubes in pressurized water reactors. Other significant corrosion problems for both reactor types are briefly reviewed

  8. Thermal analysis of biological shield of fast breeder test reactor

    International Nuclear Information System (INIS)

    Saha, D.; Sarda, V.

    1976-01-01

    A design optimisation of the biological shield of fast breeder test reactor was carried out using computer code HEATING. The effect of different heat sources, variation of coolant tube pitch circle radius, coolant temperature, angular pitch of coolant tubes and thermal conductivity of concrete on the temperature distribution within the shield has been studied. (author)

  9. Residential Energy Efficiency Potential: West Virginia

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Eric J [National Renewable Energy Laboratory (NREL), Golden, CO (United States)

    2017-11-27

    Energy used by West Virginia single-family homes that can be saved through cost-effective improvements. Prepared by Eric Wilson and Noel Merket, NREL, and Erin Boyd, U.S. Department of Energy Office of Energy Policy and Systems Analysis.

  10. Business Plan: The Virginia Space Flight Center

    Science.gov (United States)

    Reed, Billie M.

    1997-01-01

    The Virginia Commercial Space Flight Authority (VCSFA) was established on July 1, 1995 and codified at Sections 9-266.1 et seq., Code of Virginia. It is governed by an eleven person Board of Directors representing industry, state and local government and academia. VCSFA has designated the Center for Commercial Space Infrastructure as its Executive Directorate and Operating Agent. This Business Plan has been developed to provide information to prospective customers, prospective investors, state and federal government agencies, the VCSFA Board and other interested parties regarding development and operation of the Virginia Space Flight Center (VSFC) at Wallops Island. The VSFC is an initiative sponsored by VCSFA to achieve its stated objectives in the areas of economic development and education. Further, development of the VSFC is in keeping with the state's economic goals set forth in Opportunity Virginia, the strategic plan for jobs and prosperity, which are to: (1) Strengthen the rapidly growing aerospace industry in space based services including launch services, remote sensing, satellite manufacturing and telecommunications; and (2) Capitalize on intellectual and technical resources throughout the state and become a leader in the development of advanced technology businesses.

  11. Report on Homeless Families in Virginia.

    Science.gov (United States)

    Luongo, Gerardine M.; Zoller, Mary

    This report provides policymakers and advocates with information about the problems homeless families face and outlines short- and long-term solutions. Initial sections provide facts on homelessness in Virginia, an introduction, and an overview. Subsequent sections explore: (1) identification of the homeless and their characteristics; (2) causes…

  12. State Education Finance and Governance Profile: Virginia

    Science.gov (United States)

    Smith, Matthew

    2010-01-01

    This article presents the state education finance and governance profile of Virginia. The state allocates K-12 education funds on a biennial basis. Supplements and rescissions to the original funding amount are made on an annual basis in order to correct for revenue growth or deficit. The state budget currently provides more than $5 billion to…

  13. 76 FR 41411 - West Virginia Regulatory Program

    Science.gov (United States)

    2011-07-14

    ... Office of Surface Mining Reclamation and Enforcement 30 CFR Part 948 West Virginia Regulatory Program AGENCY: Office of Surface Mining Reclamation and Enforcement (OSM), Interior. ACTION: Interim rule; effective date. SUMMARY: On June 29, 2011, OSM published an interim rule approving a program amendment...

  14. Solar radiation at Parsons, West Virginia

    Science.gov (United States)

    James H. Patric; Stanley Caruso

    1978-01-01

    Twelve years of solar radiation data, measured with a Kipp-Zonen pyranometer, were recorded near Parsons, West Virginia. The data agree well with calculated values of potential and average radiation for the vicinity and are applicable to the central Appalachian region.

  15. Silvical characteristics of Virginia pine (Pinus virginiana)

    Science.gov (United States)

    Albert G., Jr. Snow

    1960-01-01

    Virginia pine has finally attained its rightful place among trees of commercial importance. It has done so in spite of being called "scrub pine" and "poverty pine" - and in spite of the term "forest weed", which has lingered long in the speech of oldtimers who remember the days of timber-plenty.

  16. KIDS COUNT in Virginia: 1997 Data Book.

    Science.gov (United States)

    Galano, Joseph; Nezlek, John B.; Wood, Lisa

    This KIDS COUNT data book examines statewide trends in the well-being of Virginia's children. The statistical portrait is based on six general areas of children's well-being: (1) healthy births; (2) children's health; (3) school success; (4) risky behavior; (5) families; and (6) community well-being. Key indicators in these six areas include the…

  17. West Virginia Dropout Study, 1985-86.

    Science.gov (United States)

    West Virginia State Dept. of Education, Charleston. Div. of General and Special Educational Development.

    Reported in this document are dropout statistics from the State of West Virginia for the school year 1985-86. This annual survey of the 55 county school systems has been conducted since the 1968-69 school year. Topics surveyed include Education Consolidation and Improvement Act (ECIA) status, exit interviews, grade at exit, month dropout left…

  18. Virginia's Value Added: A Diverse System Perspective

    Science.gov (United States)

    Harper, Vernon B., Jr.

    2009-01-01

    Every ecological, educational, or organizational system possesses a defining trait, one that affords competitive advantage. The Virginia system of higher education has consistently leaned on its institutional diversity to produce one of the most well-educated and productive bodies of learners in the nation. Given the national climate, a commitment…

  19. Retooling Teacher Preparation in West Virginia

    Science.gov (United States)

    Manchin, Gayle

    2015-01-01

    This article addresses West Virginia's public schools, and their long struggle with student achievement levels in reading and math. Levels are significantly below the national average and there are poverty-based achievement gaps within the state. In 2013, a cross section of educators and education policy leaders from a range of experiences,…

  20. Virginia, 2011 forest inventory and analysis factsheet

    Science.gov (United States)

    Anita K. Rose

    2013-01-01

    This science update is a brief look at some of the basic metrics that describe the status and trends of forest resources in Virginia. Estimates presented here are for the measurement year 2011. Information for the factsheets is updated by means of the Forest Inventory and Analysis (FIA) annualized sample design. Each year 20 percent of the sample plots (one panel) in...