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Sample records for fort st. vrain reactor

  1. Depressurization accident analyses for the Fort St. Vrain Reactor

    International Nuclear Information System (INIS)

    Paul, D.D.

    1976-01-01

    Design-basis depressurization accident analyses for the Fort St. Vrain reactor were performed using the FLODIS (Ref. 4) code. The FLODIS code models the active core, side reflector, gas annulus between the core barrel and the PCRV liner, and the PCRV cooling system. Results are presented for the Pelton circulators operating at 10,550, 8800, and 7000 rpm. Maximum temperatures of selected components are plotted as a function of time during the transient. None of the components studied exceeded the temperature at which failure or damage may occur. However, there must be sufficient mixing of the outlet gas in the lower plenum to insure the integrity of the steel liners of the steam generator inlet ducts

  2. Fort St. Vrain reactor performance and operation to full power

    International Nuclear Information System (INIS)

    Simon, W.A.; Bramblett, G.C.

    1982-01-01

    The Fort St. Vrain Nuclear Generating Station, powered by a high-temperature gas-cooled reactor (HTGR), has now been tested to full thermal power. Testing was conducted for the dual purposes of demonstrating component and system capability as a part of the rise-to-power program and determining core fluctuation/redistribution behavior under full power conditions. Both objectives were met. Full power performance of all major components and the achievement of nearly all design objectives has been verified. In addition, the tests showed that the fluctuation phenomenon has been corrected. Core region outlet temperature redistributions have been characterized, related to a physical mechanism, and shown to be inconsequential for overall plant operation

  3. Fort St. Vrain core performance

    International Nuclear Information System (INIS)

    McEachern, D.W.; Brown, J.R.; Heller, R.A.; Franek, W.J.

    1977-07-01

    The Fort St. Vrain High Temperature Gas Cooled Reactor core performance has been evaluated during the startup testing phase of the reactor operation. The reactor is graphite moderated, helium cooled, and uses coated particle fuel and on-line flow control to each of the 37 refueling regions. Principal objectives of startup testing were to determine: core and control system reactivity, radial power distribution, flow control capability, and initial fission product release. Information from the core demonstrates that Technical Specifications are being met, performance of the core and fuel is as expected, flow and reactivity control are predictable and simple for the operator to carry out

  4. Fort St. Vrain decommissioning project

    International Nuclear Information System (INIS)

    Fisher, M.

    1998-01-01

    Public Service Company of Colorado (PSCo), owner of the Fort St. Vrain nuclear generating station, achieved its final decommissioning goal on August 5, 1997 when the Nuclear Regulatory Commission terminated the Part 50 reactor license. PSCo pioneered and completed the world's first successful decommissioning of a commercial nuclear power plant after many years of operation. In August 1989, PSCo decided to permanently shutdown the reactor and proceed with its decommissioning. The decision to proceed with early dismantlement as the appropriate decommissioning method proved wise for all stake holders - present and future - by mitigating potential environmental impacts and reducing financial risks to company shareholders, customers, employees, neighboring communities and regulators. We believe that PSCo's decommissioning process set an exemplary standard for the world's nuclear industry and provided leadership, innovation, advancement and distinguished contributions to other decommissioning efforts throughout the world. (author)

  5. Fort St. Vrain circulator operating experience

    Energy Technology Data Exchange (ETDEWEB)

    Brey, H. L.

    1988-08-15

    Fort St. Vrain, on the system of Public Service Company of Colorado, is the only high-temperature gas-cooled power reactor in the United States. Four helium circulators are utilized in this plant to transfer heat from the reactor to the steam generators. These unique machines have a single stage axial flow helium compressor driven by a single stage steam turbine. A single stage water driven (pelton wheel) turbine is the back-up drive utilizing either feed water, condensate, or fire water as the driving fluid. Developmental testing of the circulators was accomplished prior to installation into Fort St. Vrain. A combined machine operating history of approximately 250,000 hours has shown these machines to be of conservative design and proven mechanical integrity. However, many problems have been encountered in operating the complex auxiliaries which are necessary for successful circulator and plant operation. It has been 15 years since initial installation of the circulators occurred at Fort St. Vrain. During this time, a number of significant issues had to be resolved dealing specifically with machine performance. These events include cavitation damage of the pelton wheels during the initial plant hot functional testing, cracks in the water turbine buckets and cervic coupling, static shutdown seal bellows failure, and, most recently, degradation of components within the steam drive assembly. Unreliable operation particularly with the circulator auxiliaries has been a focus of attention by Public Service Company of Colorado. Actions to replace or significantly modify the existing circulators and their auxiliaries are currently awaiting decisions concerning the long-term future of the Fort St. Vrain plant. (author). 10 refs, 7 figs, 2 tabs.

  6. Fort St. Vrain circulator operating experience

    International Nuclear Information System (INIS)

    Brey, H.L.

    1988-01-01

    Fort St. Vrain, on the system of Public Service Company of Colorado, is the only high-temperature gas-cooled power reactor in the United States. Four helium circulators are utilized in this plant to transfer heat from the reactor to the steam generators. These unique machines have a single stage axial flow helium compressor driven by a single stage steam turbine. A single stage water driven (pelton wheel) turbine is the back-up drive utilizing either feed water, condensate, or fire water as the driving fluid. Developmental testing of the circulators was accomplished prior to installation into Fort St. Vrain. A combined machine operating history of approximately 250,000 hours has shown these machines to be of conservative design and proven mechanical integrity. However, many problems have been encountered in operating the complex auxiliaries which are necessary for successful circulator and plant operation. It has been 15 years since initial installation of the circulators occurred at Fort St. Vrain. During this time, a number of significant issues had to be resolved dealing specifically with machine performance. These events include cavitation damage of the pelton wheels during the initial plant hot functional testing, cracks in the water turbine buckets and cervic coupling, static shutdown seal bellows failure, and, most recently, degradation of components within the steam drive assembly. Unreliable operation particularly with the circulator auxiliaries has been a focus of attention by Public Service Company of Colorado. Actions to replace or significantly modify the existing circulators and their auxiliaries are currently awaiting decisions concerning the long-term future of the Fort St. Vrain plant. (author). 10 refs, 7 figs, 2 tabs

  7. Primary coolant chemistry of the Peach Bottom and Fort St. Vrain high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Burnette, R.D.; Baldwin, N.L.

    1980-11-01

    The chemical impurities in the primary coolants of the Peach Bottom and Fort St. Vrain reactors are discussed. The impurity mixtures in the two plants were quite different because the sources of the impurities were different. In the Peach Bottom reactor, the impurities were dominated by H 2 and CH 4 , which are decomposition products of oil. In the Fort St. Vrain reactor, there were high levels of CO, CO 2 , and H 2 O. Although oil ingress at Peach Bottom created carbon deposits on virtually all surfaces, its effect on reactor operation was negligible. Slow outgassing of water from the thermal insulation at Fort St. Vrain caused delays in reactor startup. The overall graphite oxidation in both plants was negligible

  8. Primary coolant chemistry of the Peach Bottom and Fort St. Vrain high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Burnette, R.D.; Baldwin, N.L.

    1981-01-01

    The chemical impurities in the primary coolants of the Peach Bottom and Fort St. Vrain reactors are discussed. The impurity mixtures in the two plants were quite different because the sources of the impurities were different. In the Peach Bottom reactor, the impurities were dominated by H 2 and CH 4 , which are decomposition products of oil. In the Fort St. Vrain reactor, there were high levels of CO, CO 2 , and H 2 O. Although oil ingress at Peach Bottom created carbon deposits on virtually all surfaces, its effect on reactor operation was negligible. Slow outgassing of water from the thermal insulation at Fort St. Vrain caused delays in reactor startup. The overall graphite oxidation in both plants was negligible. (author)

  9. Guide to General Atomic studies of hypothetical nuclear driven accidents for the Fort St. Vrain reactor

    International Nuclear Information System (INIS)

    Wei, T.; Tobias, M.

    1974-03-01

    The work of the General Atomic Company (GAC) in preparing those portions of the Final Safety Analysis Report for the Fort St. Vrain Reactor (FSV) having to do with hypothetical nuclear driven accidents has been reviewed and a guide to this literature has been prepared. The sources for this study are the Final Safety Analysis Report itself, the Quarterly and Monthly Progress Reports, Topical Reports, and Technical Specifications. The problems considered and the methods used are outlined. An appendix gives a systematic analysis which was used as a guide in organizing the references. (U.S.)

  10. CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

    International Nuclear Information System (INIS)

    Kotas, J.F.; Stroh, K.R.

    1983-01-01

    The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

  11. Status of the Fort St. Vrain decommissioning

    International Nuclear Information System (INIS)

    Fisher, M.J.

    1990-01-01

    Fort St. Vrain is a high temperature gas cooled reactor. It has been shut down as a result of financial and technical difficulties. Fort St. Vrain has been planning for defueling and decommissioning for at least three years. The preliminary decommissioning plan, in accordance with the NRC's final rule, has been submitted and is being reviewed by the NRC. The basis of the preliminary decommissioning plan has been SAFSTOR. Public Service Company, who is the owner and operator of FSV, is scheduled to submit a proposed decommissioning plan to the NRC in the fourth quarter of 1990. PSC has gone out for bid on the decontamination and dismantlement of FSV. This paper includes the defueling schedule, the independent spent fuel storage installation status, the probability of shipping fuel to DOE, the status of the preliminary decommissioning plan submittal, the issuance of a possession only license and what are the results of obtaining this license amendment, preliminary decommissioning activities allowed prior to the approval of a proposed decommissioning plan, the preparation of a proposed decommissioning plan and the status of our decision to proceed with SAFSTOR or DECON as identified in the NRC's final decommissioning rule

  12. Fort St. Vrain defueling ampersand decommissioning considerations

    International Nuclear Information System (INIS)

    Warembourg, D.

    1994-01-01

    Fort St. Vrain Nuclear Generating Station (FSV) is one of the first commercial reactors to be decommissioned under NRC's decommissioning rule. The defueling and decommissioning of this 330 MWe High Temperature Gas Cooled Reactor (HTGR) has involved many challenges for Public Service Company of Colorado (PSC) including defueling to an Independent Spent Fuel Storage Installation (ISFSI), establishing decommissioning funding, obtaining regulatory approvals, arranging for waste disposal, and managing a large fixed price decommissioning contract. In 1990, a team comprised of the Westinghouse Corporation and Morrison Knudsen Corporation, with the Scientific Ecology Group as a major subcontractor, was contracted by PSC to perform the decommissioning under a fixed price contract. Physical work activities began in August 1992. Currently, physical dismantlement activities are about 45% complete, the project is on schedule, and is within budget

  13. Operational testing highlights of Fort St. Vrain

    International Nuclear Information System (INIS)

    Cadwell, J.J.; McEachern, D.W.; Read, J.W.; Simon, W.A.; Walker, R.F.

    1975-01-01

    The Fort St. Vrain program has progressed through construction, preoperational testing, fuel loading, initial criticality, and operational testing at power levels up to 2 percent related power. To date, all tests necessary before the rise to full power have been completed, and the rise-to-power program is expected to be resumed again in late 1975. Major plant systems, including the prestressed concrete reactor vessel and circulators, have demonstrated adequate performance. Extensive tests on the reactor core at zero power and up to 2 percent power have demonstrated the accuracy in the design predictions of such core characteristics as critical rod position, control system worths, neutron flux distributions, and temperature coefficients. Gaseous fission product release measurements to date have confirmed the extensive analytical estimates. 6 references

  14. Operational experience at Fort St. Vrain

    International Nuclear Information System (INIS)

    Bramblett, G.C.; Fisher, C.R.; Swart, F.E.

    1981-01-01

    The Fort St. Vrain (FSV) station, a 330-MW(e) single reheat steam cycle powered by a high-temperature gas-cooled reactor (HTGR), is the first HTGR to enter commercial operation. Designed and built by General Atomic Company (GA), the plant is owned and operated by Public Service Company of Colorado (PSC). Many unique design features have been incorporated into this reactor system, including high-pressure helium as the primary system coolant, a graphite-moderated prismatic block core design, fission-product-containing carbide coatings on both fissile and fertile fuel particles, steam-driven helium circulators turning on water bearings, and once-through steam generators. All of these systems are contained in a prestressed concrete reactor vessel (PCRV). Extensive testing has been conducted during the rise to power following first criticality early in 1974 to verify system design performance. During this period, the plant has operated at power levels up to 70% and produced over one billion kilowatt hours of electricity. In 1979, the first refueling was conducted in conjunction with an extensive in-core inspection, the addition of in-core instrumentation, and a planned removal of a circulator for inspection. Later in the year, a scheduled shutdown was undertaken for surveillance tests, insertion of core region constraint devices (RCDs), and other maintenance. Fort St. Vrain has encountered problems of the type that would be expected in a first-of-a-kind system. The plant is currently restricted to 70% of design power by the Nuclear Regulatory Commission (NRC) pending resolution of the core region gas outlet temperature fluctuation problem. Even so, the basic performance of the HTGR concept and all of the unique design features have been successfully demonstrated. The system has been characterized by low personnel radiation exposures, operational flexibility, and long time afforded for status evaluation and response. (author)

  15. FLODIS: a computer model to determine the flow distribution and thermal response of the Fort St. Vrain reactor

    Energy Technology Data Exchange (ETDEWEB)

    Paul, D.D.

    1976-06-01

    FLODIS is a combined heat transfer and fluid flow analysis calculation written specifically for the core of the Fort St. Vrain reactor. It is a lumped-node representation of the 37 refueling regions in the active core. Heat conduction to the coolant and in the axial direction is represented; however, the effect of conduction between refueling regions is not included. The calculation uses the specified operating conditions for the reactor at power to determine appropriate loss coefficients for the variable orifices in each refueling region. Flow distributions following reactor trip and a reduction in coolant pressure and flow are determined assuming that the orifice coefficients remain constant. Iterative techniques are used to determine the distribution of coolant flow as a function of time during the transient. Results are presented for the evaluation of the transient for the Fort St. Vrain reactor following depressurization and cooling with two circulators operating at 8000 rpm.

  16. FLODIS: a computer model to determine the flow distribution and thermal response of the Fort St. Vrain reactor

    International Nuclear Information System (INIS)

    Paul, D.D.

    1976-06-01

    FLODIS is a combined heat transfer and fluid flow analysis calculation written specifically for the core of the Fort St. Vrain reactor. It is a lumped-node representation of the 37 refueling regions in the active core. Heat conduction to the coolant and in the axial direction is represented; however, the effect of conduction between refueling regions is not included. The calculation uses the specified operating conditions for the reactor at power to determine appropriate loss coefficients for the variable orifices in each refueling region. Flow distributions following reactor trip and a reduction in coolant pressure and flow are determined assuming that the orifice coefficients remain constant. Iterative techniques are used to determine the distribution of coolant flow as a function of time during the transient. Results are presented for the evaluation of the transient for the Fort St. Vrain reactor following depressurization and cooling with two circulators operating at 8000 rpm

  17. Construction experience on PCRV liners at Fort St. Vrain

    International Nuclear Information System (INIS)

    Cliff, J.O.; Wunderlich, R.G.

    1976-01-01

    The construction of the steel liners for the Fort St. Vrain prestressed concrete reactor vessel presented many unique problems for which techniques were developed to satisfy the rigid specification requirements. The PCRV cavity liner was fabricated from 1.9cm carbon steel plate. The liners were partially fabricated by Pittsburgh-Des Moines Steel Company at their Pittsburgh manufacturing facility. The liners were then shipped by rail to within approximately five miles of the jobsite and then trucked the remaining distance. The construction techniques, dimensional control, concrete support and testing utilized on the Fort St. Vrain project are presented in detail and demonstrate the flexibility of the PCRV for field construction. (author)

  18. Safety and licensing analyses for the Fort St. Vrain HTGR

    International Nuclear Information System (INIS)

    Ball, S.J.; Conklin, J.C.; Harrington, R.M.; Cleveland, J.C.; Clapp, N.E. Jr.

    1982-01-01

    The Oak Ridge National Laboratory (ORNL) safety analysis program for the HTGR includes development and verification of system response simulation codes, and applications of these codes to specific Fort St. Vrain reactor licensing problems. Licensing studies addressed the oscillation problems and the concerns about large thermal stresses in the core support blocks during a postulated accident

  19. Operational experience at Fort St. Vrain

    Energy Technology Data Exchange (ETDEWEB)

    Bramblett, G. C.; Fisher, C. R.; Swart, F. E. [General Atomic Co., San Diego, CA (USA)

    1981-01-15

    The Fort St. Vrain (FSV) station, a 330-MW(e) single reheat steam cycle powered by a high-temperature gas-cooled reactor (HTGR), is the first HTGR to enter commercial operation. Designed and built by General Atomic Company (GA), the plant is owned and operated by Public Service Company of Colorado (PSC). Many unique design features have been incorporated into this reactor system, including high-pressure helium as the primary system coolant, a graphite-moderated prismatic block core design, fission-product-containing carbide coatings on both fissile and fertile fuel particles, steam-driven helium circulators turning on water bearings, and once-through steam generators. All of these systems are contained in a prestressed concrete reactor vessel (PCRV). Extensive testing has been conducted during the rise to power following first criticality early in 1974 to verify system design performance. During this period, the plant has operated at power levels up to 70% and produced over one billion kilowatt hours of electricity. In 1979, the first refueling was conducted in conjunction with an extensive in-core inspection, the addition of in-core instrumentation, and a planned removal of a circulator for inspection.

  20. Fort St. Vrain hot functional test results

    International Nuclear Information System (INIS)

    Phelps, R.D.

    1974-01-01

    A description is given of Fort St. Vrain hot functional tests performed to evaluate the initial nonnuclear performance of the primary coolant system and the associated effects on the various internal components of the reactor vessel and primary coolant system. The components included the twelve steam generator modules, the four helium circulators, the PCRV thermal barrier and liner coolant system, the helium purification system, and the primary and secondary closures at each of the PCRV penetrations. Additional objectives included analysis of the parallel operation of the four helium circulators and the performance of several circulator start/stop transients under various conditions of primary coolant temperature and pressure. Vibration and acoustical phenomena within the vessel were measured, recorded, and compared to theoretical analyses; a verification of reverse flow in the shutdown loop steam generator during one loop operation was performed; the PCRV was again observed for its structural response to internal pressure; and comparisons were made relative to data recorded during the initial pressure test completed in July 1971. (U.S.)

  1. Leaktightness in HTGRs - experience at Fort St. Vrain

    International Nuclear Information System (INIS)

    Neylan, A.J.; Barker, R.A.; Deardorff, A.F.

    1976-01-01

    The Fort St. Vrain Prestressed Concrete Reactor Vessel is the first utilized to contain the helium coolant of a High Temperature Gas-Cooled Reactor. Because the helium coolant contains fission products, leakage from the vessel is limited to 15 percent of vessel inventory per year. This paper describes the fabrication methods and development tests used to assure this leaktightness and the leakage test conducted to verify it. (author)

  2. Radiochemical analysis of the first plateout probe from the Fort St. Vrain high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Burnette, R.D.

    1982-06-01

    This report presents the analysis of radioactive elements on the first plateout probe from the Fort St. Vrain high-temperature gas-cooled reactor. The plateout probe is a device which samples the primary coolant for condensible fission products. Circuit inventories of individual radionuclides are estimated from the probe analysis. The analysis shows that the radioactive contamination in the primary circuit is remarkable low, with activation product concentrations much greater than that of fission products. The analysis demonstrates that the concentrations of the key fission products I-131 and Sr-90 are far below the limits allowed by the technical specification

  3. Calculations of the three-dimensional power distribution in the Fort St. Vrain reactor using UK methods and data

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, D W

    1973-04-15

    Assessments of the ability of UK methods and data developed primarily for the low enriched uranium cycle to simulate thorium cycle HTRs haye been extended to cover reactivity and power distributions in commercial size reactors. The Fort St. Vrain 330 MW(E) HTR being built in the United States by Gulf General Atomic has been chosen as a convenient object for such a study since detailed design information together with the results of GGA's own calculations have been published. The results obtained are in good agreement with those obtained by GGA and indicate that both thorium and low enriched cycle HTRs can be adequately modelled with UK data and methods.

  4. Fort St. Vrain improvement program plan. Draft final report

    International Nuclear Information System (INIS)

    1980-03-01

    The restraints are described which inhibit the Fort St. Vrain (FSV) Nuclear Power Station, a high temperature gas cooled reactor (HTGR) plant, from achieving full power operation with high availability. The actions necessary to overcome these restraints are outlined. The restraints originated from problems in both hardware related and institutional areas. The report summarizes what has been accomplished, what is currently being done, and what should be done to resolve the problems

  5. Reflector dowel strength test, Fort St. Vrain

    International Nuclear Information System (INIS)

    Doll, D.W.

    1975-01-01

    The strength of the 44.45 mm (1.75 in.) diameter Fort St. Vrain (FSV) reflector dowel for loads directed radially inward toward the center of the element was measured. For a statically applied load, the strength exceeded 5783 N (1300 lb) in direct shear. This strength remained after load cycling 100 times to 4448 N (1000 lb), 10 times to 4893 N (1100 lb), 10 times to 5338 N (1200 lb), and two times to 5783 N (1300 lb). Typically, the deflection to ultimate failure was approximately 1.0 mm (0.04 in.). At about 3316 N (750 lb) and 0.20 mm (0.008 in.) deflection, one of the webs between the dowel and a coolant hole cracked, apparently redistributing the load. No further failure occurred up to the ultimate load of 5783+ N (1300+ lb)

  6. Dynamic computer simulation of the Fort St. Vrain steam turbines

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1983-01-01

    A computer simulation is described for the dynamic response of the Fort St. Vrain nuclear reactor regenerative intermediate- and low-pressure steam turbines. The fundamental computer-modeling assumptions for the turbines and feedwater heaters are developed. A turbine heat balance specifying steam and feedwater conditions at a given generator load and the volumes of the feedwater heaters are all that are necessary as descriptive input parameters. Actual plant data for a generator load reduction from 100 to 50% power (which occurred as part of a plant transient on November 9, 1981) are compared with computer-generated predictions, with reasonably good agreement

  7. Construction, testing, and initial operation of Fort St. Vrain PCRV

    International Nuclear Information System (INIS)

    Ople, F.S. Jr.; Neylan, A.J.

    1975-01-01

    The Fort St. Vrain (FSV) Nuclear Generating Station is the first station in the USA to use a prestressed concrete reactor vessel (PCRV). The PCRV was designed and constructed by General Atomic. Construction of the PCRV was completed in 1970; the pressure and leak tests were completed in 1971. The structural behavior of the PCRV has been monitored by installed instrumentation since start of construction. The highlights of the actual construction, testing, and initial operation of the PCRV, including a comparison of structural behavior, where possible, between observed data and analytical predictions. (U.S.)

  8. ALARA and decommissioning: The Fort St. Vrain experience

    Energy Technology Data Exchange (ETDEWEB)

    Borst, T.; Niehoff, M. [Public Service Co. of Colorado, Platteville, CO (United States); Zachary, M. [Scientific Ecology Group, Platteville, CO (United States)

    1995-03-01

    The Fort St. Vrain Nuclear Generating Station, the first and only commercial High Temperature Gas Cooled Reactor to operate in the United States, completed initial fuel loading in late 1973 and initial startup in early 1974. Due to a series of non-nuclear technical problems, Fort St. Vrain never operated consistently, attaining a lifetime capacity factor of slightly less than 15%. In August of 1989, the decision was made to permanently shut down the plant due to control rod drive and steam generator ring header failures. Public Service Company of Colorado elected to proceed with early dismantlement (DECON) as opposed to SAFSTOR on the bases of perceived societal benefits, rad waste, and exposure considerations, regulatory uncertainties associated with SAFSTOR, and cost. The decommissioning of Fort St. Vrain began in August of 1992, and is scheduled to be completed in early 1996. Decommissioning is being conducted by a team consisting of Westinghouse, MK-Ferguson, and Scientific Ecology Group. Public Service Company of Colorado as the licensee provides contract management and oversight of contractor functions. An aggressive program to maintain project radiation exposures As Low As Reasonably Achievable (ALARA) has been established, with the following program elements: temporary and permanent shielding contamination control; mockup training; engineering controls; worker awareness; integrated work package reviews communication; special instrumentation; video camera usage; robotics application; and project committees. To date, worker exposures have been less than project estimates. from the start of the project through Februrary of 1994, total exposure has been 98.666 person-rem, compared to the project estimate of 433 person-rem and goal of 347 person-rem. The presentation will discuss the site characterization efforts, the radiological performance indicator program, and the final site release survey plans.

  9. ORNL's NRC-sponsored HTGR safety and licensing analysis activities for Fort St. Vrain and advanced reactors

    International Nuclear Information System (INIS)

    Ball, S.J.; Cleveland, J.C.; Harrington, R.M.

    1985-01-01

    The ORNL safety analysis program for the HTGR was established in 1974 to provide technical assistance to the USNRC on licensing questions for both Fort St. Vrain and advanced plant concepts. The emphasis has been on development of major component and system dynamic simulation codes, and use of these codes to analyze specific licensing-related scenarios. The program has also emphasized code verification, using Fort St. Vrain data where applicable, and comparing results with industry-generated codes. By the use of model and parameter adjustment routines, safety-significant uncertainties have been identified. A major part of the analysis work has been done for the Fort St. Vrain HTGR, and has included analyses of FSAR accident scenario re-evaluations, the core block oscillation problem, core support thermal stress questions, technical specification upgrade review, and TMI action plan applicability studies. The large, 2240-MW(t) cogeneration lead plant design was analyzed in a multi-laboratory cooperative effort to estimate fission product source terms from postulated severe accidents

  10. Preplaced aggregate concrete application on Fort St. Vrain PCRV construction

    International Nuclear Information System (INIS)

    Ople, F.S. Jr.

    1976-01-01

    Two distinct concreting methods were employed in the construction of the prestressed concrete reactor vessel (PCRV) of the Fort St. Vrain (FSV) Nuclear Generating Station, a 330 MW(e) High Temperature Gas-Cooled Reactor installation near Denver, Colorado. Preplaced aggregate concrete (PAC) techniques were employed in the PCRV bottom head and the core support floor; conventional job-mixed concrete was used in the PCRV sidewall and top head regions. This paper describes the successful application of PAC techniques utilized primarily in solving construction difficulties associated with confined and heavily congested regions of the PCRV. The PAC technique consists of placing coarse aggregate inside the forms, followed by injection of grout under pressure through embedded pipes to fill the interstices in the aggregate mass. Details of the PAC construction method including grout mix development, grouting equipment, grout pipe layout, grouting sequence, grout level monitoring, concrete temperature control, and pre-construction mockups are described. (author)

  11. Fort St. Vrain graphite site mechanical separation concept selection

    International Nuclear Information System (INIS)

    Berry, S.M.

    1993-09-01

    One of the alternatives to the disposal of the Fort St. Vrain (FSV) reactor spent nuclear fuel involves the separation of the fuel rods composed of compacts from the graphite fuel block assembly. After the separation of these two components, the empty graphite fuel blocks would be disposed of as a low level waste (provided the appropriate requirements are met) and the fuel compacts would be treated as high level waste material. This report deals with the mechanical separation aspects concerning physical disassembly of the FSV graphite fuel element into the empty graphite fuel blocks and fuel compacts. This report recommends that a drilling technique is the preferred choice for accessing the, fuel channel holes and that each hole is drilled separately. This report does not cover any techniques or methods to separate the triso fuel particles from the graphite matrix of the fuel compacts

  12. Neutron flux distribution measurement in the Fort St. Vrain initial core (results of Fort St. Vrain start-up test A-7)

    International Nuclear Information System (INIS)

    Marshall, A.C.; Brown, J.R.

    1975-01-01

    A description is given of a test to measure the axial flux distribution at several radial locations in the Fort St. Vrain core representing unrodded, rodded, and partially rodded regions. The measurements were intended to verify the calculational accuracy of the three-dimensional calculational model used to compute axial power distributions for the Fort St. Vrain core. (U.S.)

  13. Fission product behavior in the Peach Bottom and Fort St. Vrain HTGRs

    International Nuclear Information System (INIS)

    Hanson, D.L.; Baldwin, N.L.; Strong, D.E.

    1980-11-01

    Actual operating data from Peach Bottom and Fort St. Vrain were compared with code predictions to assess the validity of the methods used to predict the behavior of fission products in the primary coolant circuit. For both reactors the measured circuit activities were significantly below design values, and the observations generally verify the codes used for large HTGR design

  14. Analysis and evaluation of recent operational experience from the Fort St. Vrain HTGR

    International Nuclear Information System (INIS)

    Moses, D.L.; Lanning, W.D.

    1985-05-01

    The Fort St. Vrain operating experience to be discussed here includes notable safety-related events which have occurred since late 1981 when ORNL was first contracted to provide technical assistance to AEOD. Earlier Fort St. Vrain operating experience through the time of successful full-power testing in November 1981 has been summarized by the licensee and the reactor vendor, GA Technologies, Inc. (GA), in papers presented at several different forums during 1982. In addition, extensive and very useful detailed evaluations of preoperational and startup testing and of the rise-to-power operating experience through completion of the first refueling outage in August 1979 have been compiled into a series of reports under the sponsorship of the Electric Power Research Institute (EPRI). Finally, the US Department of Energy's Fort St. Vrain Improvement Plan provides a summary of the major operational limits which have affected the plant since start-up. The events discussed here are categorized based on the major systems affected, namely, (1) primary system and reactor vessel, (2) electrical systems, and (3) the reactor building. In all cases to be discussed, the lessons to be learned are vigilance and prevention. These lessons translate into the need for the recognition and control of unexpected situations and of their potential for branching effects. At Fort St. Vrain, these lessons are found in the effects of moisture ingress, in the challenges experienced to the supply of essential electrical power, and in controlling the environment of the reactor building. 13 refs

  15. Fort Saint Vrain operational experience

    International Nuclear Information System (INIS)

    Fuller, C.H.

    1989-01-01

    Fort St. Vrain (FSV), on the system of the Public Service Company of Colorado, is the only high temperature gas-cooled (HTGR) power reactor in the United States. The plant features a helium-cooled reactor with a uranium-thorium fuel cycle. The paper describes the experience made during its operation. (author). 2 refs, 4 figs, 2 tabs

  16. Test and evaluation of the Fort St. Vrain dew point moisture monitor system

    International Nuclear Information System (INIS)

    Block, G.A.; Del Bene, J.V. Jr.; Gitterman, M.; Hastings, G.A.; Hawkins, W.M.; Hinz, R.F.; McCue, D.E.; Swanson, L.L.; Vavrina, J.; Zwetzig, G.B.

    1975-01-01

    Descriptions are given of the Fort St. Vrain Dew Point Moisture Monitor (DPMM) System; the bases for the DPMM system response time requirements for safety related functions at the required reactor operating conditions; the results and evaluation of recent testing which measured the performance of the current system at simulated operating conditions; predicted response times for reactor power operation from 0 to 100 percent and a modification to provide improved response times for low-load and plant start-up conditions

  17. The Fort St. Vrain high temperature gas-cooled reactor. III

    International Nuclear Information System (INIS)

    Olson, H.G.; Brey, H.L.

    1979-01-01

    The helium circulator auxiliary system provides buffer helium and bearing water for the reactor's four circulators with two nearly identical auxiliary loops serving the two circulators of a primary coolant loop. A series of drains removes the water and helium for separation and recycle. Loss of buffer helium's function as a dynamic seal has resulted in inleakage of bearing water into the primary coolant and outleakage of primary coolant into the auxiliary system. Inleakage of water also has occurred due to inadvertent pressurization of the bearing cavity with the static shutdown seal set. Satisfactory performance of the normal, backup and emergency bearing water systems has been accomplished after numerous component additions and modifications. Frequent circulator trips have occurred. Most of these have involved the delicate sensors that measure buffer helium differential pressure. Transients in one loop have communicated to the other loop through common components. Total separation of the auxiliary loops will occur after the planned installation of those components that currently service both loops. (Auth.)

  18. Equipment for nondestructive evaluation of the strength of the Fort St. Vrain core-support blocks

    International Nuclear Information System (INIS)

    Morgan, W.C.; Prince, J.M.; Posakony, G.J.

    1982-09-01

    A novel sweep-frequency eddy current instrument has been constructed for measuring density-depth profiles in oxidized graphite. Development work on additional parts of the instrumentation package, that was to be tested in the Fort St. Vrain High Temperature Gas-Cooled Reactor, has been temporarily halted. This report documents the work which has been accomplished to date and presents the current status of the equipment development effort

  19. Nondestructive examination of 51 fuel and reflector elements from Fort St. Vrain Core Segment 1

    International Nuclear Information System (INIS)

    Miller, C.M.; Saurwein, J.J.

    1980-12-01

    Fifty-one fuel and reflector elements irradiated in core segment 1 of the Fort St. Vrain High-Temperature Gas-Cooled Reactor (HTGR) were inspected dimensionally and visually in the Hot Service Facility at Fort St. Vrain in July 1979. Time- and volume-averaged graphite temperatures for the examined fuel elements ranged from approx. 400 0 to 750 0 C. Fast neutron fluences varied from approx. 0.3 x 10 25 n/m 2 to 1.0 x 10 25 n/m 2 (E > 29 fJ)/sub HTGR/. Nearly all of the examined elements shrank in both axial and radial dimensions. The measured data were compared with strain and bow predictions obtained from SURVEY/STRESS, a computer code that employs viscoelastic beam theory to calculate stresses and deformations in HTGR fuel elements

  20. Fort St. Vrain fuel-handling system RAM analysis

    International Nuclear Information System (INIS)

    Azizi, S.M.; Berg, G.E.; Burton, J.H.; Durand, R.E.; Larson, E.M.; Pepe, D.J.; Rutherford, P.D.; Novachek, F.J.

    1989-01-01

    Public Service of Company of Colorado (PSC) is planning to decommission its Fort St. Vrain plant in 1990. This requires removal of 1,500 separate assemblies from the core. With the low historical availability of the fuel-handling system (FHS), defueling time was estimated at 36 months. With plant expenses of approximately $1.6 million per month during defueling, this would mean a schedule cost of $58 million. With their contractor, Rockwell International, PSC embarked on a reliability, availability, and maintainability (RAM) analysis to reduce projected defueling time. Key elements included (a) estimating availability of the FHS using a limited historical record, (b) assessing the defueling critical path, and (c) proposing and evaluating design/operational improvements. The most cost-effective improvements are being implemented and are expected to provide a reduction of >18 months in schedule and a net savings of $20 to 25 million. The paper describes the FHS design and operation, major problems associated with fuel-handling operations, and results and recommendations

  1. Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility

    International Nuclear Information System (INIS)

    1992-07-01

    Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility

  2. Failure of Fort St. Vrain 347SS control rod drive cables

    International Nuclear Information System (INIS)

    Hellner, R.L.; Thurgood, B.E.

    1990-01-01

    This paper reports on Fort St. Vrain (FSV) which is a high temperature gas cooled reactor. During a scheduled surveillance exercise, one of the control rod drives failed to operate properly. It was found that one of the 347 austenitic stainless cables had failed at several locations and the other had a broken strand. Metallurgical examination determined that the cables failed due to chloride stress corrosion cracking. An investigation into the source of chlorides determined that materials within the core could release chlorides either by water leaching or heat up. To prevent future failures, all the stainless control cables were replaced with cables fabricated from inconel 625

  3. Technical evaluation report of the Fort St. Vrain final draft upgraded technical specifications

    International Nuclear Information System (INIS)

    Kimura, C.Y.

    1989-01-01

    This report is a technical evaluation of the final draft of the Fort St. Vrain (FSV) Upgraded Technical Specifications (UT/S) as issued by Public Service of Colorado (PSC) on May 27, 1988 with subsequent supplemental updates issued on June 15, 1988 and August 5, 1988. It has been compared for consistency, and safety conservatism with the Fort St. Vrain (FSV) Updated Final Safety Analysis Report (FSAR), the FSV Safety Evaluation Report (SER), the Facility Operating License, DPR-34, and all amendments to the Facility Operating License issued as of June 1, 1988, and Appendix A to the Operating License DPR-34, Technical Specifications. Because of the age of the plant, no supplements to the Fort St. Vrain SER have been issued since the original SER was not issued as a WASH or a NUREG report. This made it necessary to review all amendments to the Facility Operating License since they would contain the safety evaluations done to support changes to the Facility Operating License. The upgraded Fort St. Vrain Technical Specifications were also broadly compared with the latest Westinghouse Standard Technical Specifications (WSTS) to assure that what was proposed for Fort St. Vrain was consistent with the latest NRC staff practices for standard technical specifications

  4. Assessment of effects of Fort St. Vrain HTGR primary coolant on Alloy 800. Final report

    International Nuclear Information System (INIS)

    Trester, P.W.; Johnson, W.R.; Simnad, M.T.; Burnette, R.D.; Roberts, D.I.

    1982-08-01

    A comprehensive review was conducted of primary helium coolant chemistry data, based on current and past operating histories of helium-cooled, high-temperature reactors (HTGRs), including the Fort St. Vrain (FSV) HTGR. A reference observed FSV reactor coolant environment was identified. Further, a slightly drier expected FSV coolant chemistry was predicted for reactor operation at 100% of full power. The expected environment was compared with helium test environments used in the US, United Kingdom, Germany, France, and Japan. Based on a comprehensive review and analysis of mechanical property data reported for Alloy 800 tested in controlled-impurity helium environments (and in air when appropriate for comparison), an assessment was made of the effect of FSV expected helium chemistry on material properties of alloy 800, with emphasis on design properties of the Alloy 800 material utilized in the FSV steam generators

  5. Conceptual design report for the mechanical disassembly of Fort St. Vrain fuel elements

    Energy Technology Data Exchange (ETDEWEB)

    Lord, D.L. [Westinghouse Idaho Nuclear Co., Inc., Idaho Falls, ID (United States); Wadsworth, D.C.; Sekot, J.P.; Skinner, K.L. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1993-04-01

    A conceptual design study was prepared that: (1) reviewed the operations necessary to perform the mechanical disassembly of Fort St. Vrain fuel elements; (2) contained a description and survey of equipment capable of performing the necessary functions; and (3) performed a tradeoff study for determining the preferred concepts and equipment specifications. A preferred system was recommended and engineering specifications for this system were developed.

  6. Conceptual design report for the mechanical disassembly of Fort St. Vrain fuel elements

    International Nuclear Information System (INIS)

    Lord, D.L.; Wadsworth, D.C.; Sekot, J.P.; Skinner, K.L.

    1993-04-01

    A conceptual design study was prepared that: (1) reviewed the operations necessary to perform the mechanical disassembly of Fort St. Vrain fuel elements; (2) contained a description and survey of equipment capable of performing the necessary functions; and (3) performed a tradeoff study for determining the preferred concepts and equipment specifications. A preferred system was recommended and engineering specifications for this system were developed

  7. Investigations of postulated accident sequences for the Fort St. Vrain HTGR

    International Nuclear Information System (INIS)

    Ball, S.J.; Cleveland, J.C.; Conklin, J.C.; Hatta, M.; Sanders, J.P.

    1978-01-01

    The systems analysis capability of the ORNL HTGR Safety analysis research program includes a family of computer codes: an overall plant NSSS simulation (ORTAP), and detailed component codes for investigating core neutronic accidents (CORTAP), shutdown emergency-cooling accidents via a 3-dimensional core model (ORECA), and once-through steam generator transients (BLAST). The component codes can either be run independently or in the overall NSSS code. Verification efforts have consisted primarily of using existing Fort St. Vrain reactor dynamics data to compare against code predictions. Comparisons of core thermal conditions made for reactor scrams from power levels between 30 and 50% showed good agreement. An optimization program was used to rationalize the difference between the predicted and measured refueling region outlet temperatures, and, in general, excellent agreement was attained by adjustment of models and parameters within their uncertainty ranges. However, more work is required to establish a unique and valid set of models

  8. Developmental assessment of the Fort St. Vrain version of the Composite HTGR Analysis Program (CHAP-2)

    International Nuclear Information System (INIS)

    Stroh, K.R.

    1980-01-01

    The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain (FSV) version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are being used to partially verify the component modeling and dynamic smulation techniques used to predict plant response to postulated accident sequences

  9. Nondestructive examination of 54 fuel and reflector elements from Fort St. Vrain core segment 2

    International Nuclear Information System (INIS)

    Saurwein, J.J.

    1982-10-01

    Fifty-four fuel and reflector elements irradiated in core segment 2 of the Fort St. Vrain high-temperature gas-cooled reactor (HTGR) were nondestructively examined. The time- and volume-averaged graphite irradiation temperatures for the elements ranged from approx. 350 0 to 750 0 C. The element-averaged fast neutron fluences ranged from approx. 0.2 to 1.6 x 10 25 n/m 2 (E > 29 fJ)/sub HTGR/. The elements, except for two fuel elements in which single localizeed cracks developed during irradiation, were in excellent condition. No evidence was observed of significant graphite oxidation or mechanical interaction beween elements. The cracks in the two elements did not affect their performance or handling. These elements were, otherwise, in excellent condition. Nearly all elements shrank in both the axial and radial directions, but the dimensional changes were relatively small

  10. Fort St. Vrain high temperature gas-cooled reactor. Pt. 12. The dew point moisture monitor testing program

    Energy Technology Data Exchange (ETDEWEB)

    Olson, H.G. (Colorado State Univ., Fort Collins (USA). Dept. of Mechanical Engineering); Brey, H.L. (Public Service Co. of Colorado, Denver (USA)); Swart, F.E. (Gas-Cooled Reactor Associates, La Jolla, CA (USA)); Forbis, J.M. (Storage Technology Corp., Louisville, CO (USA))

    1982-09-01

    Moisture ingress into the core volume could cause damaging reactions with the moderator-reflector graphite and burnable poison, therefore a dew point moisture monitoring system has been developed with the basic design criteria that a plant protective system trip is signaled after the system detects high primary coolant helium moisture levels and that the system is able to correctly identify which of two steam generator loops is leaking. Modifications to the sample supplies to the monitors were necessary to reduce the system's unsatisfactory response time at lower reactor power levels.

  11. Developmental assessment of the Fort St. Vrain version of the composite HTGR analysis program (CHAP-2)

    International Nuclear Information System (INIS)

    Stroh, K.R.

    1981-01-01

    The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are being used to partially verify the component modeling and dynamic simulation techniques used to predict plant response to postulated accident sequences. Results of these preliminary validation efforts are presented showing good agreement between code output and plant data for the portions of the code that have been tested. Plans for further development and assessment as well as application of the validated code are discussed. (author)

  12. Thermal protection system for the concrete core support floor at Fort St. Vrain

    International Nuclear Information System (INIS)

    Jones, H.; Hedgecock, P.D.

    1976-01-01

    A unique feature of the Fort St. Vrain HTGR is its steel jacketed concrete core support floor. The construction of this floor generally resembles that of the prestressed concrete reactor vessel, but its location immediately below the core hot gas outlets generates some particularly severe thermal protection requirements. A thermal barrier is used over the entire outer surface of the floor and in the twelve hot gas ducts which convey the primary coolant through the floor to the steam generators. A cooling water system of square tubes welded to the inside of the steel jacket is used to remove that heat which does pass through the thermal barrier and to maintain the concrete at acceptable temperatures. The design approach to the floor itself and to the thermal barriers and cooling system will be described, but the main emphasis of the paper will be on the total experience gained during construction and pre-operational testing. A particular problem experienced during construction was leakage from some cooling tubes, after their embedment in concrete. The solution to that problem was to develop a method for injecting catalyzed epoxy into the leaking tube. This method, which has general usefulness for in-service repairs, will be described. (author)

  13. Fission product behaviour in the Peach Bottom and Fort St. Vrain HTGRs

    International Nuclear Information System (INIS)

    Hanson, D.L.; Baldwin, N.L.; Strong, D.E.

    1981-01-01

    Actual operating data from the Peach Bottom (PB) and Fort St. Vrain (FSV) High-Temperature Gas-Cooled Reactors (HTGRs) have been compared with code predictions to assess the validity of the methods used to predict the behaviour of fission products in the primary coolant circuit. For both reactors the measured circuit activities were significantly below design values, and the observations generally verify the codes used for large HTGR design. The PB primary circuit after seven years of operation was exceptionally clean. A fuel element purge system virtually eliminated the release of fission gases into the primary coolant circuit. Extensive examinations at end-of-life revealed that only Cs and trace amounts of Sr had plated out in the circuit. Their plateout distributions were in excellent agreement with PAD code predictions. Most of the deposited activity was associated with carbonaceous surface films which resulted from occasional small inleakages of lubricating oil. Primary circuit activities in FSV during the first cycle were also very low. Noble gas activity was about 1% of the design limit; and the circulating iodines were at least one order of magnitude below the limit, although the measurement uncertainties are significant. The plateout per pass of the iodine isotopes increased with decreasing half-life (the value for I-131 is about 1% per pass) as predicted with the PADLOC code. Gamma scanning of two helium circulators indicated very low plateout activities. Iodine-131 was the principal fission product observed, along with small amounts of Cs-134, Cs-137, and Ba/La-140. (author)

  14. Postirradiation examination and evaluation of Fort St. Vrain fuel element 1-0743

    International Nuclear Information System (INIS)

    Saurwein, J.J.; Miller, C.M.; Young, C.A.

    1981-05-01

    Fort St. Vrain (FSV) fuel element 1-0743 was irradiated in core location 17.04.F.06 from July 3, 1976 until February 1, 1979. The element experienced an average fast neutron exposure of about 0.95 x 10 25 n/m 2 (E > 29 fJ)/sub HTGR/, a time-and-volume-averaged fuel temperature in the vicinity of 680 0 C, fissile and fertile particle burnups of approximately 6.2% and 0.3%, respectively, and a total burnup of 12,210 MWd/tonne. The postirradiation examination revealed that the element was in excellent condition. No cracks were observed on any of the element surfaces. The structural integrity of the fuel rods was good. No evidence of mechanical interaction between the fuel rods and fuel body was observed. Calculated irradiation parameters obtained with HTGR design codes were compared with measured data. Radial and axial power distributions, irradiation temperatures, neutron fluences, and fuel burnups were in good agreement with measurements. Calculated fuel rod strains were about a factor of three greater than were observed

  15. Uranium and thorium loadings determined by chemical and nondestructive methods in HTGR fuel rods for the Fort St. Vrain Early Validation Irradiation Experiment

    International Nuclear Information System (INIS)

    Angelini, P.; Rushton, J.E.

    1979-01-01

    The Fort St. Vrain Early Validation Irradiation Experiment is an irradiation test of reference and of improved High-Temperature Gas-Cooled Reactor fuels in the Fort St. Vrain Reactor. The irradiation test includes fuel rods fabricated at ORNL on an engineering scale fuel rod molding machine. Fuel rods were nondestructively assayed for 235 U content by a technique based on the detection of prompt-fission neutrons induced by thermal-neutron interrogation and were later chemically assayed by using the modified Davies Gray potentiometric titration method. The chemical analysis of the thorium content was determined by a volumetric titration method. The chemical assay method for uranium was evaluated and the results from the as-molded fuel rods agree with those from: (1) large samples of Triso-coated fissile particles, (2) physical mixtures of the three particle types, and (3) standard solutions to within 0.05%. Standard fuel rods were fabricated in order to evaluate and calibrate the nondestructive assay device. The agreement of the results from calibration methods was within 0.6%. The precision of the nondestructive assay device was established as approximately 0.6% by repeated measurements of standard rods. The precision was comparable to that estimated by Poisson statistics. A relative difference of 0.77 to 1.5% was found between the nondestructive and chemical determinations on the reactor grade fuel rods

  16. Comparison of predicted and measured fission product behaviour in the Fort St. Vrain HTGR during the first three cycles of operation

    International Nuclear Information System (INIS)

    Hanson, D.L.; Jovanovic, V.; Burnette, R.D.

    1985-01-01

    The 330 MW(e) Fort St. Vrain (M) High Temperature Gas-Cooled Reactor (HTGR) is fueled with (Th,U)C 2 /ThC 2 TRISO-coated fuel particles contained in prismatic graphite fuel elements. Fission product release from the reactor core has been monitored during the first three cycles of operation. In order to assess the validity of the design methods used to predict fission product source terms for HTGRs, fission product release from the reactor core has been predicted by the reference design methods and compared with reactor surveillance measurements and with the results of postirradiation examination (PIE) of spent FSV fuel elements. Overall, the predictive methods have been shown to be conservative: the predicted fission gas release at the end of Cycle 3 is about five times higher than observed. The dominant source of fission gas release is as-manufactured, heavy-metal contamination; in-service failure of the coated fuel particles appears to be negligible, which is consistent with the PIE of spent fuel elements removed during the first two refuelings. The predicted releases of fission metals are insignificant compared to the release and subsequent decay of their gaseous precursors, which is consistent with plateout probe measurements. (author)

  17. Technical and regulatory review of the Rover nuclear fuel process for use on Fort St. Vrain fuel

    International Nuclear Information System (INIS)

    Hertzler, T.

    1993-02-01

    This report describes the results of an analysis for processing and final disposal of Fort St. Vrain (FSV) irradiated fuel in Rover-type equipment or technologies. This analysis includes an evaluation of the current Rover equipment status and the applicability of this technology in processing FSV fuel. The analyses are based on the physical characteristics of the FSV fuel and processing capabilities of the Rover equipment. Alternate FSV fuel disposal options are also considered including fuel-rod removal from the block, disposal of the empty block, or disposal of the entire fuel-containing block. The results of these analyses document that the current Rover hardware is not operable for any purpose, and any effort to restart this hardware will require extensive modifications and re-evaluation. However, various aspects of the Rover technology, such as the successful fluid-bed burner design, can be applied with modification to FSV fuel processing. The current regulatory climate and technical knowledge are not adequately defined to allow a complete analysis and conclusion with respect to the disposal of intact fuel blocks with or without the fuel rods removed. The primary unknowns include the various aspects of fuel-rod removal from the block, concentration of radionuclides remaining in the graphite block after rod removal, and acceptability of carbon in the form of graphite in a high level waste repository

  18. Initial Start-Up and Testing of the Fort St. Vrain HTGR – Lessons Learned which May Be Useful for the HTR-PM

    International Nuclear Information System (INIS)

    Brey, H.L.

    2014-01-01

    Although the activities presented in this paper occurred 40 years ago, there are many observations and lessons associated with Fort St. Vrain (FSV) which may be beneficial in support of the start-up, testing and licensing of the HTR-PM. This report includes a review of the FSV NPP design including an overview of the requirements and testing program utilized to bring the plant from initial start-up to full power. A sampling of the test results as well as a comparison of the plant design characteristics to actual values achieved at 100% power along with selected overall experiences gained through operation of this plant is also included. (author)

  19. Fort St. Vrain Nuclear Generating Station. Semiannual operations report, January--June 1975

    International Nuclear Information System (INIS)

    1975-01-01

    The reactor was in start-up testing during this period. Information is presented concerning fuel and equipment performance; changes in facility design; procedures; testing; staff; licensed operator training; surveillance; power generation; maintenance; shutdowns, primary coolant chemistry; radiation exposure; fission product release; solid wastes; and radioactive effluents

  20. A reactivity accidents simulation of the Fort Saint Vrain HTGR

    International Nuclear Information System (INIS)

    Fainer, Gerson

    1980-01-01

    A reactivity accidents analysis of the Fort Saint Vrain HTGR was made. The following accidents were analysed 1) A rod pair withdrawal accident during normal operation, 2) A rod pair ejection accident, 3) A rod pair withdrawal accident during startup operations at source levels and 4) Multiple rod pair withdrawal accident. All the simulations were performed by using the BLOOST-6 nuclear code The steady state reactor operation results obtained with the code were consistent with the design reactor data. The numerical analysis showed that all accidents - except the first one - cause particle failure. (author)

  1. Calculation of Void in the Fort Saint Vrain Material

    Energy Technology Data Exchange (ETDEWEB)

    Potter, David Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Taylor, Craig Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Coons, James Elmer [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-11

    The percent void of the Fort Saint Vrain (FSV) material is estimated to be 21.1% based on the volume of the gap at the top of the drums, the volume of the coolant channels in the FSV fuel element, and the volume of the fuel handling channel in the FSV fuel element.

  2. Nondestructive evaluation of the oxidation and strength of the Fort Saint Vrain HTGR support block

    International Nuclear Information System (INIS)

    Tingey, G.L.; Posakony, G.J.; Morgan, W.C.; Prince, J.M.; Hill, R.W.; Lessor, D.L.

    1982-04-01

    Non-destructive detection of changes in the strength of graphite support structures in a HTGR appears to be feasible using sonic velocity measurements where access for through transmission is possible. Therefore, future HTGR designs should consider providing such access. Where access is not available, strength changes can be correlated with oxidation profiles in the support member. These oxidation profiles can be determined non-destructively by a combination of eddy current measurements to detect near surface oxidation and sonic backscattering measurements designed to determine oxidation in depth. The Fort Saint Vrain reactor provides an operating reactor to test the applicability of the eddy current and sonic backscattering techniques for determination of oxidation in a support block. Furthermore, such tests in Fort Saint Vrain will supply base line data which will be useful in assuring an adequate strength of the support structure for the lifetime of the reactor. Equipment is, therefore, being developed for tests to be conducted during the next major refueling of the reactor

  3. Geochemical Interactions in failed Co-Disposal Waste Packages for N Reactor and Ft. St. Vrain Spent Fuel and the Melt and Dilute Waste Form

    International Nuclear Information System (INIS)

    Arthur, S.E.; McNeish, J.

    2002-01-01

    The objective of this scientific analysis is to calculate the long-term geochemical behavior in a failed co-disposal waste package (WP) containing U. S. Department of Energy (DOE) spent nuclear fuel (SNF) and high level waste (HLW) glass. This analysis was prepared according to a Technical Work Plan (BSC 2002). Specifically the scope of these calculations is to determine: (1) The geochemical characteristics of the fluids inside the WP after breach, including the corrosion/dissolution of the initial WP configuration; (2) The transport of radionuclides of concern to performance assessment out of the degraded WP by infiltrating water; and (3) The range of parameter variation for additional laboratory and numerical evaluations. This analysis is limited to three SNF groups, uranium (U)/thorium (Th) carbide SNF (Group 5), U metal SNF (Group 7), and aluminum(Al)-based fuels (Group 9). Group 5 is represented by Ft. St. Vrain (FSV) U/Th carbide SNF, Group 7 is represented by N-Reactor U metal SNF, and Group 9 is represented by the Melt and Dilute (MandD) waste form developed from Al-based SNF. The DOE (2001a, Appendix A) describes all of these fuels. Table 1 shows the groups of DOE SNF, the representative SNF for each group, and the metric tons of heavy metal (MTHM) of SNF in each group

  4. Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

    International Nuclear Information System (INIS)

    Radulescu, H.

    2001-01-01

    There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report

  5. Evaluation of Codisposal Viability for TH/U Carbide (Fort Saint Vrain HTGR) DOE-Owned Fuel

    Energy Technology Data Exchange (ETDEWEB)

    H. radulescu

    2001-09-28

    There are more than 250 forms of US Department of Energy (DOE)-owned spent nuclear fuel (SNF). Due to the variety of the spent nuclear fuel, the National Spent Nuclear Fuel Program has designated nine representative fuel groups for disposal criticality analyses based on fuel matrix, primary fissile isotope, and enrichment. The Fort Saint Vrain reactor (FSVR) SNF has been designated as the representative fuel for the Th/U carbide fuel group. The FSVR SNF consists of small particles (spheres of the order of 0.5-mm diameter) of thorium carbide or thorium and high-enriched uranium carbide mixture, coated with multiple, thin layers of pyrolytic carbon and silicon carbide, which serve as miniature pressure vessels to contain fission products and the U/Th carbide matrix. The coated particles are bound in a carbonized matrix, which forms fuel rods or ''compacts'' that are loaded into large hexagonal graphite prisms. The graphite prisms (or blocks) are the physical forms that are handled in reactor loading and unloading operations, and which will be loaded into the DOE standardized SNF canisters. The results of the analyses performed will be used to develop waste acceptance criteria. The items that are important to criticality control are identified based on the analysis needs and result sensitivities. Prior to acceptance to fuel from the Th/U carbide fuel group for disposal, the important items for the fuel types that are being considered for disposal under the Th/U carbide fuel group must be demonstrated to satisfy the conditions determined in this report.

  6. Parametric study of the low-enriched uranium integrated Fort-Saint-Vrain element; comparative evaluation with the interacting tubular element

    International Nuclear Information System (INIS)

    Cerles, J.M.; Carvallo, G.; Vallepin, C.

    1971-11-01

    This paper presents a study of the influence of the different geometric and neutronic parameters on the calculation of the cycle with low-enriched uranium in a Fort-Saint-Vrain type brick. The study is divided in two parts: a stage of physics, essentially neutronics; an economical part where the costs are taken into account. At the level of studies of neutronics and costs, a parallel comparison is developed between the brick Fort-Saint-Vrain and the interacting tubular element, and even thorium. 6 refs. 29 figs [fr

  7. RAPID Assessment of Extreme Reservoir Sedimentation Resulting from the September 2013 Flood, North St. Vrain Creek, CO

    Science.gov (United States)

    Rathburn, S. L.; McElroy, B. J.; Wohl, E.; Sutfin, N. A.; Huson, K.

    2014-12-01

    During mid-September 2013, approximately 360 mm of precipitation fell in the headwaters of the North St. Vrain drainage basin, Front Range, CO. Debris flows on steep hillslopes and extensive flooding along North St. Vrain Creek resulted in extreme sedimentation within Ralph Price Reservoir, municipal water supply for the City of Longmont. The event allows comparison of historical sedimentation with that of an unusually large flood because 1) no reservoir flushing has been conducted since dam construction, 2) reservoir stratigraphy chronicles uninterrupted delta deposition, and 3) this is the only on-channel reservoir with unimpeded, natural sediment flux from the Continental Divide to the mountain front in a basin with no significant historic flow modifications and land use impacts. Assessing the flood-related sedimentation prior to any dredging activities included coring the reservoir delta, a bathymetric survey of the delta, resistivity and ground penetrating radar surveys of the subaerial inlet deposit, and surveying tributary deposits. Over the 44-year life of the reservoir, two-thirds of the delta sedimentation is attributed to extreme discharges from the September 2013 storm. Total storm-derived reservoir sedimentation is approximately 275,000 m3, with 81% of that within the gravel-dominated inlet and 17% in the delta. Volumes of deposition within reservoir tributary inlets is negatively correlated with contributing area, possibly due to a lack of storage in these small basins (1-5 km2). Flood-related reservoir sedimentation will be compared to other research quantifying volumes from slope failures evident on post-storm lidar. Analysis of delta core samples will quantify organic carbon flux associated with the extreme discharge and develop a chronology of flood and fire disturbances for North St. Vrain basin. Applications of similar techniques are planned for two older Front Range reservoirs affected by the September flooding to fill knowledge gaps about

  8. The present state of development and the future of the high-temperature reactor in the United States of America

    International Nuclear Information System (INIS)

    Simon, W.A.; Chi, H.W.

    1982-01-01

    The American prototype high-temperature reactor at Fort St. Vrain has been operating successfully for years. To date it has produced more than 3.000.000.000 kilowatt hours of electricity and a short while ago was cleared for operation at full load. Operating experience justifies expectations that the combined cycle HTR plant of 2240 MW thermal output favoured by the US Government and industry will offer significant economic advantages. (orig.) [de

  9. Aquatic Communities and Selected Water Chemistry in St. Vrain Creek near the City of Longmont, Colorado, Wastewater-Treatment Plant, 2005 and 2006

    Science.gov (United States)

    Zuellig, Robert E.; Sprague, Lori A.; Collins, Jim A.; Cox, Oliver N.

    2007-01-01

    In 2005, the U.S. Geological Survey and the City of Longmont, Colo., began a study to document chemical characteristics of St. Vrain Creek that had previously been unavailable either due to high cost of analysis or lack of analytical capability. Stream samples were collected at seven sites on St. Vrain Creek during the spring of 2005 and 2006 for analysis of wastewater compounds. A Lagrangian-sampling design was followed during each sampling event, and time-of-travel studies were conducted just prior to each sampling event to determine appropriate sampling times for the synoptic. In addition, semipermeable membrane devices, passive samplers that concentrate hydrophobic organic chemicals, were installed at six sites during the spring of 2005 and 2006 for approximately 4 weeks. After retrieval, contaminant residues concentrated in the semipermeable membrane devices were recovered and used in a toxicity assay that provided a screen for aryl hydrocarbon receptor type compounds, including polycyclic aromatic hydrocarbons, polychlorinated biphenyls, dioxins, and furans. In addition, the U.S. Geological Survey summarized information on macroinvertebrate and fish communities known from St. Vrain Creek dating back to the early 1900s in order to assess their utility in evaluating wastewater-treatment plant upgrades and habitat improvement projects. Unfortunately, because of inconsistencies in data collection these data cannot be used as intended; however, they are useful for understanding to some degree gross patterns in fish species distribution, but less so for macroinvertebrates.

  10. Requirements of, and operating experience with, gas analyses on high temperature reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1982-06-01

    Impurities in the helium coolant of the primary coolant circuit of HTGR's are mainly due to ingress of air or water, occasionally oil. Typical concentrations are given of H 2 O, H 2 , CO 2 , CO, N 2 , CH 4 and Ar in the AVR, Dragon, Peach Bottom and Fort St. Vrain reactors. A characteristic is presented of measuring devices for measuring non-active impurities in helium; measuring methods are described and a list is given of required and actual detection limits. Also given are concentrations of solid fission and activation products and tritium in the primary circuit of the AVR reactor

  11. Public Service Company, Fort St. Vrain Station, Petition For Objection to Issuance of Operating Permit

    Science.gov (United States)

    This document may be of assistance in applying the Title V air operating permit regulations. This document is part of the Title V Petition Database available at www2.epa.gov/title-v-operating-permits/title-v-petition-database. Some documents in the database are a scanned or retyped version of a paper photocopy of the original. Although we have taken considerable effort to quality assure the documents, some may contain typographical errors. Contact the office that issued the document if you need a copy of the original.

  12. Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1986-08-01

    The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory

  13. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  14. Medium-size high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Peinado, C.O.; Koutz, S.L.

    1980-08-01

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 760 0 C (1400 0 F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics [a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant] and engineered safety features

  15. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Trauger, D.B.

    1976-01-01

    Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing

  16. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    Energy Technology Data Exchange (ETDEWEB)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched {sup 235}U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched {sup 235}U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing.

  17. Options for treating high-temperature gas-cooled reactor fuel for repository disposal

    International Nuclear Information System (INIS)

    Lotts, A.L.; Bond, W.D.; Forsberg, C.W.; Glass, R.W.; Harrington, F.E.; Micheals, G.E.; Notz, K.J.; Wymer, R.G.

    1992-02-01

    This report describes the options that can reasonably be considered for disposal of high-temperature gas-cooled reactor (HTGR) fuel in a repository. The options include whole-block disposal, disposal with removal of graphite (either mechanically or by burning), and reprocessing of spent fuel to separate the fuel and fission products. The report summarizes what is known about the options without extensively projecting or analyzing actual performance of waste forms in a repository. The report also summarizes the processes involved in convert spent HTGR fuel into the various waste forms and projects relative schedules and costs for deployment of the various options. Fort St. Vrain Reactor fuel, which utilizes highly-enriched 235 U (plus thorium) and is contained in a prismatic graphite block geometry, was used as the baseline for evaluation, but the major conclusions would not be significantly different for low- or medium-enriched 235 U (without thorium) or for the German pebble-bed fuel. Future US HTGRs will be based on the Fort St. Vrain (FSV) fuel form. The whole block appears to be a satisfactory waste form for disposal in a repository and may perform better than light-water reactor (LWR) spent fuel. From the standpoint of process cost and schedule (not considering repository cost or value of fuel that might be recycled), the options are ranked as follows in order of increased cost and longer schedule to perform the option: (1) whole block, (2a) physical separation, (2b) chemical separation, and (3) complete chemical processing

  18. Fuel-Cycle and Nuclear Material Disposition Issues Associated with High-Temperature Gas Reactors

    International Nuclear Information System (INIS)

    Shropshire, D.E.; Herring, J.S.

    2004-01-01

    The objective of this paper is to facilitate a better understanding of the fuel-cycle and nuclear material disposition issues associated with high-temperature gas reactors (HTGRs). This paper reviews the nuclear fuel cycles supporting early and present day gas reactors, and identifies challenges for the advanced fuel cycles and waste management systems supporting the next generation of HTGRs, including the Very High Temperature Reactor, which is under development in the Generation IV Program. The earliest gas-cooled reactors were the carbon dioxide (CO2)-cooled reactors. Historical experience is available from over 1,000 reactor-years of operation from 52 electricity-generating, CO2-cooled reactor plants that were placed in operation worldwide. Following the CO2 reactor development, seven HTGR plants were built and operated. The HTGR came about from the combination of helium coolant and graphite moderator. Helium was used instead of air or CO2 as the coolant. The helium gas has a significant technical base due to the experience gained in the United States from the 40-MWe Peach Bottom and 330-MWe Fort St. Vrain reactors designed by General Atomics. Germany also built and operated the 15-MWe Arbeitsgemeinschaft Versuchsreaktor (AVR) and the 300-MWe Thorium High-Temperature Reactor (THTR) power plants. The AVR, THTR, Peach Bottom and Fort St. Vrain all used fuel containing thorium in various forms (i.e., carbides, oxides, thorium particles) and mixtures with highly enriched uranium. The operational experience gained from these early gas reactors can be applied to the next generation of nuclear power systems. HTGR systems are being developed in South Africa, China, Japan, the United States, and Russia. Elements of the HTGR system evaluated included fuel demands on uranium ore mining and milling, conversion, enrichment services, and fuel fabrication; fuel management in-core; spent fuel characteristics affecting fuel recycling and refabrication, fuel handling, interim

  19. Flood-inundation maps for the St. Marys River at Fort Wayne, Indiana

    Science.gov (United States)

    Menke, Chad D.; Kim, Moon H.; Fowler, Kathleen K.

    2012-01-01

    Digital flood-inundation maps for a 9-mile reach of the St. Marys River that extends from South Anthony Boulevard to Main Street at Fort Wayne, Indiana, were created by the U.S. Geological Survey (USGS) in cooperation with the City of Fort Wayne. The inundation maps, which can be accessed through the USGS Flood Inundation Mapping Science Web site, depict estimates of the areal extent of flooding corresponding to selected water levels (stages) at the USGS streamgage 04182000 St. Marys River near Fort Wayne, Ind. Current conditions at the USGS streamgages in Indiana may be obtained from the National Water Information System: Web Interface. In addition, the information has been provided to the National Weather Service (NWS) for incorporation into their Advanced Hydrologic Prediction Service (AHPS) flood warning system. The NWS forecasts flood hydrographs at many places that are often collocated at USGS streamgages. That forecasted peak-stage information, also available on the Internet, may be used in conjunction with the maps developed in this study to show predicted areas of flood inundation. In this study, water-surface profiles were simulated for the stream reach by means of a hydraulic one-dimensional step-backwater model. The model was calibrated using the most current stage-discharge relation at the USGS streamgage 04182000 St. Marys River near Fort Wayne, Ind. The hydraulic model was then used to simulate 11 water-surface profiles for flood stages at 1-ft intervals referenced to the streamgage datum and ranging from bankfull to approximately the highest recorded water level at the streamgage. The simulated water-surface profiles were then combined with a geographic information system digital elevation model (derived from Light Detection and Ranging (LiDAR) data) in order to delineate the area flooded at each water level. A flood inundation map was generated for each water-surface profile stage (11 maps in all) so that for any given flood stage users will be

  20. Reactor G1: high power experiments; Experiences a forte puissance

    Energy Technology Data Exchange (ETDEWEB)

    Laage, F de; Teste du Baillet, A; Veyssiere, A; Wanner, G [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Retel, H [Societe Rateau, D.E.A. (France)

    1957-07-01

    The experiments carried out in the starting-up programme of the reactor G1 comprised a series of tests at high power, which allowed the following points to be studied: 1- Effect of poisoning by Xenon (absolute value, evolution). 2- Temperature coefficients of the uranium and graphite for a temperature distribution corresponding to heating by fission. 3- Effect of the pressure (due to the coiling system) on the reactivity. 4- Calibration of the security rods as a function of their position in the pile (1). 5- Temperature distribution of the graphite, the sheathing, the uranium and the air leaving the canals, in a pile running normally at high power. 6- Neutron flux distribution in a pile running normally at high power. 7- Determination of the power by nuclear and thermodynamic methods. These experiments have been carried out under two very different pile conditions. From the 1. to the 15. of August 1956, a series of power increases, followed by periods of stabilisation, were induced in a pile containing uranium only, in 457 canals, amounting to about 34 tons of fuel. A knowledge of the efficiency of the control rods in such a pile has made it possible to measure with good accuracy the principal effects at high temperatures, that is, to deal with points 1, 2, 3, 5. Flux charts giving information on the variations of the material Laplacian and extrapolation lengths in the reflector have been drawn up. Finally the thermodynamic power has been measured under good conditions, in spite of some installation difficulties. On September 16, the pile had its final charge of 100 tons. All the canals were loaded, 1,234 with uranium and 53 (i.e. exactly 4 per cent of the total number) with thorium uniformly distributed in a square lattice of 100 cm side. Since technical difficulties prevented the calibration of the control rods, the measurements were limited to the determination of the thermodynamic power and the temperature distributions (points 5 and 7). This report will

  1. A new small modular high-temperature gas-cooled reactor plant concept based on proven technology

    International Nuclear Information System (INIS)

    McDonald, C.F.; Goodjohn, A.J.

    1982-01-01

    Based on the established and proven high-temperature gas-cooled reactor (HTGR) technologies from the Peach Bottom 1 and Fort St. Vrain utility-operated units, a new small modular HTGR reactor is currently being evaluated. The basic nuclear reactor heat source, with a prismatic core, is being designed so that the decay heat can be removed by passive means (i.e., natural circulation). Although this concept is still in the preconceptual design stage, emphasis is being placed on establishing an inherently safe or benign concept which, when engineered, will have acceptable capital cost and power generation economics. The proposed new HTGR concept has a variety of applications, including electrical power generation, cogeneration, and high-temperature process heat. This paper discusses the simplest application, i.e., a steam Rankine cycle electrical power generating version. The gas-cooled modular reactor concepts presented are based on a graphite moderated prismatic core of low-power density (i.e., 4.1 W/cm 3 ) with a thermal rating of 250 MW(t). With the potential for inherently safe characteristics, a new small reactor could be sited close to industrial and urban areas to provide electrical power and thermal heating needs (i.e., district and space heating). Incorporating a multiplicity of small modular units to provide a larger power output is also discussed. The potential for a small, inherently safe HTGR reactor concept is highlighted

  2. Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

    International Nuclear Information System (INIS)

    McDonald, C.F.; Nichols, M.K.; Kaufman, J.S.

    1988-01-01

    Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs

  3. Public Service Company, Fort St. Vrain Station, Order Granting in Part and Denying in Part Petition for Objection to Title V Permit

    Science.gov (United States)

    This document may be of assistance in applying the Title V air operating permit regulations. This document is part of the Title V Petition Database available at www2.epa.gov/title-v-operating-permits/title-v-petition-database. Some documents in the database are a scanned or retyped version of a paper photocopy of the original. Although we have taken considerable effort to quality assure the documents, some may contain typographical errors. Contact the office that issued the document if you need a copy of the original.

  4. Loading of fuel and reflector elements in the Fort St. Vrain initial core (results of start-up test A-1)

    International Nuclear Information System (INIS)

    Marshall, A.C.; Brown, J.R.

    1974-01-01

    R A description is given of the experimental equipment and techniques used in the fuel and reflector loading. The analysis methods are described and test data are compared with predicted results. (U.S.)

  5. Initial Startup and Testing of the Fort St. Vrain HTGR - Lessons Learned which May Be Useful to the HTR-PM

    International Nuclear Information System (INIS)

    Brey, Larry H.

    2014-01-01

    Lessons Learned: Although the HTR-PM and FSV incorporate significant differences in their designs, there are lessons to be learned that are applicable to both plants. This is especially important for key systems that incorporate first-of-a-kind equipment. Basically, these lessons are just an application of common sense. • Complexity Breeds Unavailability. Incorporate system/components that are ruggedly simple in design with a history of reliable operation and minimal maintenance. • Assure Strong Expertise and Funding for this First HTR-PM. Quite likely, the successful startup and operation of this plant will require a level of support considerably greater than a typical nuclear plant. • Be Very Attentive to the Design Aspects of first-of-a-kind Components in the Class 1, Safety-Related Portions of the Plant. For example; a generic metallurgical failure could easily lead to a very long plant shutdown in order to redesign the failed component, re-license, manufacture, install and test prior to plant resuming plant operation. • Where Possible, Test all Key Systems/Components Prior to Installation using Actual Plant Configuration & Operating Characteristics This will help assure operational capability prior to application of nuclear heat. • Never Attempt to Start an Innovative Nuclear Power Plant Without First Having the Proper Regulatory Guides and Criteria in Place. FSV was licensed as a Research Facility. There was no Standard Review Plan or Regulatory Guides in place for the NRC (or PSC) to utilize in regulating this HTGR. • Do Not Be Reluctant to Incorporate a Generous Over-Build Capability into Systems/Components. It is significantly easier to design extra margin into the original compressors, pumps and motors than to be required to backfit into larger units after plant start-up. • Assure All Safety Documents Reflect the Actual Capability of the Plant to Respond to Accidents Described in the Safety Analysis. FSV was limited to 82% power during the final two years of operation.

  6. Development of a surveillance robot for dimensional and visual inspection of fuel and reflector elements from the Fort St. Vrain HTGR

    International Nuclear Information System (INIS)

    Wallroth, C.F.; Marsh, N.I.; Miller, C.M.; Saurwein, J.J.; Smith, T.L.

    1979-11-01

    A robotic device has been developed for dimensional and visual inspection of irradiated HTGR core components. The robot consists of a rotary table and a two-finger probe, driven by stepping motors, and four remotely controlled television cameras. Automated operation is accomplished via minicomputer control. A total of 51 irradiated fuel and reflector elements were inspected at a fraction of the time and cost required for conventional methods

  7. HTGR safety research concerns at NRC

    International Nuclear Information System (INIS)

    Minogue, R.B.

    1982-01-01

    A general discussion of HTGR technical and safety-related problems is given. The broad areas of current research programs specific to the Fort St. Vrain reactor and applicable to HTGR technology are summarized

  8. ORTURB, HTGR Steam Turbine Dynamic for FSV Reactor

    International Nuclear Information System (INIS)

    Conklin, J.C.

    2001-01-01

    1 - Description of program or function: ORTURB was written specifically to calculate the dynamic behavior of the Fort St. Vrain (FSV) High- Temperature Gas-Cooled Reactor (HTGR) steam turbines. The program is divided into three main parts: the driver subroutine; turbine subroutines to calculate the pressure-flow balance of the high-, intermediate-, and low-pressure turbines; and feedwater heater subroutines. 2 - Method of solution: The program uses a relationship derived for ideal gas flow in an iterative fashion that minimizes computational time to determine the pressure and flow in the FSV steam turbines as a function of plant transient operating conditions. An important computer modeling characteristic, unique to FSV, is that the high-pressure turbine exhaust steam is used to drive the reactor core coolant circulators prior to entering the reheater. A feedwater heater dynamic simulation model utilizing seven state variables for each of the five heaters is included in the ORTURB computer simulation of the regenerative Rankine cycle steam turbines. The seven temperature differential equations are solved at each time- step using a matrix exponential method. 3 - Restrictions on the complexity of the problem: The turbine shaft is assumed to rotate at a constant (rated) speed of 3600 rpm. Energy and mass storage of steam in the high-, intermediate-, and low-pressure turbines is assumed to be negligible. These limitations exclude the use of ORTURB during a turbine transient such as startup from zero power or very low turbine flows

  9. Reactor technology. Progress report, July-September 1980

    International Nuclear Information System (INIS)

    Breslow, M.

    1980-12-01

    Progress in the Space Power Advanced Reactor (SPAR) Program includes indications that revision of the BeO reflector configuration can reduce system weight. Observed boiling limit restrictions on the performance of the annular-wick core heat pipe have accelerated transition to the development of the target-design arterial heat pipe. Successful bends of core heat pipes have been made with sodium as the mandrel material. With the phasing out of the GCFR Program, work on the Low Power Safety Experiments Program is now concentrated on completion of the third 37-rod Full Length Subgroup test. In the Reactor Safety/Structural Analysis area, effort on the Category I Structures Program is toward developing an experimental test plan focusing on a specific structural design. Buckling experiments on thin-walled cylindrical shells with circular cutouts are reported. Results of a three-dimensional analysis of thermal stresses in the Fort St. Vrain core support block are presented. Materials investigations and operation of a molybdenum-core SiC heat pipe are reported. Entrainment limits for gravity-assisted heat pipes and heat pipe configurations for application to energy conservation are being investigated. The new solution critical assembly, SHEBA, was completed. Godiva IV was temporarily relocated at TA-15. Influence of scattered radiations in the test vault on InRad measurements was determined from detector scans of the vault produced by 252 Cf neutron and 137 Cs gamma sources

  10. Development of a detailed core flow analysis code for prismatic fuel reactors

    International Nuclear Information System (INIS)

    Bennett, R.G.

    1990-01-01

    The detailed analysis of the core flow distribution in prismatic fuel reactors is of interest for modular high-temperature gas-cooled reactor (MHTGR) design and safety analyses. Such analyses involve the steady-state flow of helium through highly cross-connected flow paths in and around the prismatic fuel elements. Several computer codes have been developed for this purpose. However, since they are proprietary codes, they are not generally available for independent MHTGR design confirmation. The previously developed codes do not consider the exchange or diversion of flow between individual bypass gaps with much detail. Such a capability could be important in the analysis of potential fuel block motion, such as occurred in the Fort St. Vrain reactor, or for the analysis of the conditions around a flow blockage or misloaded fuel block. This work develops a computer code with fairly general-purpose capabilities for modeling the flow in regions of prismatic fuel cores. The code, called BYPASS solves a finite difference control volume formulation of the compressible, steady-state fluid flow in highly cross-connected flow paths typical of the MHTGR

  11. CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Cleveland, J.C.

    1977-01-01

    CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered region of the reactor core. The space independent neutron kinetics equations are coupled to the heat transfer equations through a rapidly converging iterative technique. The code has the capability to determine conservative fuel, moderator, and coolant temperatures in the ''hot'' fuel region. For transients involving a reactor trip, the core heat generation rate is determined from an expression for decay heat following a scram. Nonlinear effects introduced by temperature dependent fuel, moderator, and coolant properties are included in the model. CORTAP predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in CORTAP

  12. High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Beck; L. F. Pincock

    2011-04-01

    The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

  13. A safety assessment of the use of graphite in nuclear reactors licensed by the US NRC

    International Nuclear Information System (INIS)

    Schweitzer, D.G.; Gurinsky, D.H.; Kaplan, E.; Sastre, C.

    1987-09-01

    This report reviews existing literature and knowledge on graphite burning and on stored energy accumulation and releases in order to assess what role, if any, a stored energy release can have in initiating or contributing to hypothetical graphite burning scenarios in research reactors. It also addresses the question of graphite ignition and self-sustained combustion in the event of a loss-of-coolant accident (LOCA). The conditions necessary to initiate and maintain graphite burning are summarized and discussed. From analyses of existing information it is concluded that only stored energy accumulations and releases below the burning temperature (650 0 C) are pertinent. After reviewing the existing knowledge on stored energy it is possible to show that stored energy releases do not occur spontaneously, and that the maximum stored energy that can be released from any reactor containing graphite is a very small fraction of the energy produced during the first few minutes of a burning incident. The conclusions from these analyses are that the potential to initiate or maintain a graphite burning incident is essentially independent of the stored energy in the graphite, and depends on other factors that are unique for these reactors, research reactors, and for Fort St. Vrain. In order to have self-sustained rapid graphite oxidation in any of these reactors, certain necessary conditions of geometry, temperature, oxygen supply, reaction product removal, and a favorable heat balance must be maintained. There is no new evidence associated with either the Windscale Accident or the Chernobyl Accident that indicates a credible potential for a graphite burning accident in any of the reactors considered in this review

  14. Steam generator materials performance in high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Chafey, J.E.; Roberts, D.I.

    1980-11-01

    This paper reviews the materials technology aspects of steam generators for HTGRs which feature a graphite-moderated, uranium-thorium, all-ceramic core and utilizes high-pressure helium as the primary coolant. The steam generators are exposed to gas-side temperatures approaching 760 0 C and produce superheated steam at 538 0 C and 16.5 MPa (2400 psi). The prototype Peach Bottom I 40-MW(e) HTGR was operated for 1349 EFPD over 7 years. Examination after decommissioning of the U-tube steam generators and other components showed the steam generators to be in very satisfactory condition. The 330-MW(e) Fort St. Vrain HTGR, now in the final stages of startup, has achieved 70% power and generated more than 1.5 x 10 6 MWh of electricity. The steam generators in this reactor are once-through units of helical configuration, requiring a number of new materials factors including creep-fatigue and water chemistry control. Current designs of larger HTGRs also feature steam generators of helical once-through design. Materials issues that are important in these designs include detailed consideration of time-dependent behavior of both base metals and welds, as required by current American Society of Mechanical Engineers (ASME) Code rules, evaluation of bimetallic weld behavior, evaluation of the properties of large forgings, etc

  15. Archeological Testing at Fort St. Leon (16PL35), Plaquemines Parish, Louisiana.

    Science.gov (United States)

    1983-05-01

    Bend which is nebulous at best. 3. Test excavations downriver from the American fort (BHT 12, 13, 14, 15) (Figure 24b) encountered disturbed strata...unidentifiable 3 Unit 1, 2.07-1.82 m Plastron fragments (pond slider turtle ) 3 Unit 1, 2.02-1.82 m Plastron and carapace fragments (pond slider turtle ) 2 Unit...trench walls less than 1 m below the surface. The same is true for the pond slider turtle bones from Units 1 and 2. 160 . . . .. . . . . . . o Nutria

  16. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

  17. Latest developments in prestressed concrete vessels for gas-cooled reactors

    International Nuclear Information System (INIS)

    Ople, F.S. Jr.

    1979-01-01

    This paper is an update of the design development of prestressed concrete vessels, commonly referred to as 'PCRVs' starting with the first single-cavity PCRV for the Fort St. Vrain Nuclear Generating Station to the latest multi-cavity PCRV configurations being utilized as the primary reactor vessels for both the High Temperature Gas-Cooled Reactor (HTGR) and the Gas-Cooled Fast Breeder Reactor (GCFR) in the U.S.A. The complexity of PCRV design varies not only due to the type of vessel configuration (single versus multi-cavity) but also on the application to the specific type of reactor concept. PCRV technology as applied to the Steam Cycle HTGR is fairly well established; however, some significant technical complexities are associated with PCRV design for the Gas Turbine HTGR and the GCFR. For the Gas Turbine HTGR, for instance, the fluid dynamics of the turbo-machinery cause multi-pressure conditions to exist in various portions of the power conversion loops during operation. This condition complicates the design approach and the proof test specification for the PCRV. The geometric configuration of the multi-cavity PCRV is also more complex due to the introduction of large horizontal cylindrical cavities (housing the turbo/machines for the Gas Turbine HTGR and circulators for the GCFR) in addition to the vertical cylindrical cavities for the core and heat exchangers. Because of this complex geometry, it becomes difficult to achieve an optimum prestressing arrangement for the PCRV. Other novel features of the multi-cavity PCRV resulting from the continuing design optimization effort are the incorporation of an asymmetric (offset core) configuration and the use of large vessel cavity/penetration concrete closures directly held down by prestressing tendons for both economic and safety reasons. (orig.)

  18. Life extension of the St. Lucie unit 1 reactor vessel

    International Nuclear Information System (INIS)

    Rowan, G.A.; Sun, J.B.; Mott, S.L.

    1991-01-01

    In late 1989, Florida Power and Light Company (FP and L) established the policy that St. Lucie unit 1 should not be prevented from achieving a 60-yr operating life by reactor vessel embrittlement. A 60-yr operating life means that the plant would be allowed to operate until the year 2036, which is 20 years beyond the current license expiration date of 2016. Since modifications to the reactor vessel and its components are projected to be expensive, the desire of FP and L management was to achieve this lifetime extension through the use of fuel management and proven technology. The following limitations were placed on any acceptable method for achieving this lifetime extension capability: low fuel cycle cost; low impact on safety parameters; very little or no operations impact; and use of normal reactor materials. A task team was formed along with the Advanced Nuclear Fuels Company (ANF) to develop a vessel-life extension program

  19. Informal support networks of low-income senior women living alone: evidence from Fort St. John, BC.

    Science.gov (United States)

    Ryser, Laura; Halseth, Greg

    2011-01-01

    Within the context of an aging Canadian rural and small-town landscape, there is a growing trend of low-income senior women living alone. While there is a perception that rural seniors have well-developed social networks to meet their daily needs, some research suggests that economic and social restructuring processes have impacted the stability of seniors' support networks in small places. While much of the research on seniors' informal networks focuses upon small towns in decline, booming resource economies can also produce challenges for low-income senior women living alone due to both a higher cost of living and the retrenchment of government and service supports. Under such circumstances, an absence of informal supports can impact seniors' health and quality of life and may lead to premature institutionalization. Drawing upon a household survey in Fort St. John, British Columbia, we explore informal supports used by low-income senior women living alone in this different context of the Canadian landscape. Our findings indicate that these women not only have a support network that is comparable to other groups, but that they are also more likely to draw upon such supports to meet their independent-living needs. These women rely heavily on family support, however, and greater efforts are needed to diversify both their formal and informal sources of support as small family networks can quickly become overwhelmed.

  20. Aerial radiological survey of the area surrounding the St. Lucie Power Plant, Fort Pierce, Florida

    International Nuclear Information System (INIS)

    Feimster, E.L.

    1979-06-01

    An airborne radiological survey of an 1100 km 2 area surrounding the St. Lucie Power Plant was conducted 1 to 8 March 1977. Detected radioisotopes and their associated gamma ray exposure rates were consistent with that expected from the normal background emitters. Count rates observed at 150 m altitude are converted to equivalent exposure rates at 1 m above the ground and are presented in the form of an isopleth map. Ground exposure rates measured with small portable instruments and soil sample analysis agreed with the airborne data. Geological data are presented in an isopleth map of rock and soil types. Also included is a brief description of the vegetation and terrain surrounding the site

  1. Korean nuclear reactor strategy for the early 21st century

    International Nuclear Information System (INIS)

    Lee, Byong Whi; Shin, Young Kyun

    1991-01-01

    The system analysis for Korean nuclear power reactor option is made on the basis of reliability, cost minimization, finite uranium resource availability and nuclear engineering manpower supply constraints. The reference reactor scenarios are developed considering the future electricity demand, nuclear share, current nuclear power plant standardization program and manufacturing capacity. The levelized power generation cost, uranium requirement and nuclear engineering professionals demand are estimated for each reference reactor scenarios and nuclear fuel cycle options from the year 1990 up to the year 2030. Based on the outcomes of the analysis, uranium resource utilization, reliability and nuclear engineering manpower requirements are sensitive to the nuclear reactor strategy and associated fuel cycle whereas the system cost is not. APWR, CANDU: FBR strategy is to be the best option for Korea. However, APWR, CANDU: Passive Safe Reactor (PSR) vFBR strategy should be also considered as a contingency for growing national concerns on nuclear safety and public acceptance deterioration in the future. FBR development and establishment of related fuel cycle should be started as soon as possible considering the uranium shortage anticipated between 2007 and 2032. It should be noted that the increasing use of nuclear energy to minimize the greenhouse effects in the early 21st century would accelerate the uranium resource depletion. The study also concludes that the current level of nuclear engineering professionals employment is not sufficient until 2010 for the establishment of nuclear infrastructure. (Author)

  2. The gas-cooled high temperature reactor. Perspectives, problems and programmes

    International Nuclear Information System (INIS)

    Beckurts, K.H.; Engelmann, P.; Erb, D.E.

    1977-01-01

    For nearly 20 years extensive research and development programmes on helium-cooled high temperature reactors (HTR) have been carried out in several countries of the world. As a result of the long-standing efforts, satisfactory solutions have been found for many of the basic problems of this new reactor system, particularly in the field of high temperature fuels and materials technology. Three small experimental plants have been operated successfully over extended periods of time. Prototype steam-cycle plants of 300MW(e) are under way at Fort St. Vrain (full-power operation scheduled for 1977) and at Schmehausen (scheduled for 1979). Major delays have occurred in the construction and commissioning of these plants for various reasons but do not reveal specific problems of the HTR. Commercial market introduction of the steam-cycle electricity generating system has been attempted, but the first approach has not been successful. Major efforts both by governments and industry are now required to ensure a successful second approach. To reach competitivity with established nuclear power systems and to take full advantage of the fuel conservation potential of the HTR requires the implementation of the closed thorium fuel cycle on a commercial scale. While some key steps of this cycle have been implemented on a laboratory scale, progress towards a prototype recycling facility has been slow. Closing the thorium fuel cycle represents a major challenge and can only be achieved in a close international collaboration. The paper discusses the world-wide status and potential of HTR technology and reviews the major international development programmes. (author)

  3. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  4. Adapting the deep burn in-core fuel management strategy for the gas turbine - modular helium reactor to a uranium-thorium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)]. E-mail: alby@neutron.kth.se; Gudowski, Waclaw [Department of Nuclear and Reactor Physics, Royal Institute of Technology, Roslagstullsbacken 21, S-10691, Stockholm (Sweden)

    2005-11-15

    In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine - modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium-thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: {sup 235}U, which represents the 20% of the fresh uranium, {sup 233}U, which is produced by the transmutation of fertile {sup 232}Th, and {sup 239}Pu, which is produced by the transmutation of fertile {sup 238}U. In order to compensate the depletion of {sup 235}U with the breeding of {sup 233}U and {sup 239}Pu, the quantity of fertile nuclides must be much larger than that one of {sup 235}U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of {sup 235}U. At the same time, the amount of {sup 235}U must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k {sub eff} and mass

  5. Adapting the deep burn in-core fuel management strategy for the gas turbine - modular helium reactor to a uranium-thorium fuel

    International Nuclear Information System (INIS)

    Talamo, Alberto; Gudowski, Waclaw

    2005-01-01

    In 1966, Philadelphia Electric has put into operation the Peach Bottom I nuclear reactor, it was the first high temperature gas reactor (HTGR); the pioneering of the helium-cooled and graphite-moderated power reactors continued with the Fort St. Vrain and THTR reactors, which operated until 1989. The experience on HTGRs lead General Atomics to design the gas turbine - modular helium reactor (GT-MHR), which adapts the previous HTGRs to the generation IV of nuclear reactors. One of the major benefits of the GT-MHR is the ability to work on the most different types of fuels: light water reactors waste, military plutonium, MOX and thorium. In this work, we focused on the last type of fuel and we propose a mixture of 40% thorium and 60% uranium. In a uranium-thorium fuel, three fissile isotopes mainly sustain the criticality of the reactor: 235 U, which represents the 20% of the fresh uranium, 233 U, which is produced by the transmutation of fertile 232 Th, and 239 Pu, which is produced by the transmutation of fertile 238 U. In order to compensate the depletion of 235 U with the breeding of 233 U and 239 Pu, the quantity of fertile nuclides must be much larger than that one of 235 U because of the small capture cross-section of the fertile nuclides, in the thermal neutron energy range, compared to that one of 235 U. At the same time, the amount of 235 U must be large enough to set the criticality condition of the reactor. The simultaneous satisfaction of the two above constrains induces the necessity to load the reactor with a huge mass of fuel; that is accomplished by equipping the fuel pins with the JAERI TRISO particles. We start the operation of the reactor with loading fresh fuel into all the three rings of the GT-MHR and after 810 days we initiate a refueling and shuffling schedule that, in 9 irradiation periods, approaches the equilibrium of the fuel composition. The analysis of the k eff and mass evolution, reaction rates, neutron flux and spectrum at the

  6. Potentialities of high temperature reactors (HTR)

    International Nuclear Information System (INIS)

    Hittner, D.

    2001-01-01

    This articles reviews the assets of high temperature reactors concerning the amount of radioactive wastes produced. 2 factors favors HTR-type reactors: high thermal efficiency and high burn-ups. The high thermal efficiency is due to the high temperature of the coolant, in the case of the GT-MHR project (a cooperation between General Atomic, Minatom, Framatome, and Fuji Electric) designed to burn Russian military plutonium, the expected yield will be 47% with an outlet helium temperature of 850 Celsius degrees. The high temperature of the coolant favors a lot of uses of the heat generated by the reactor: urban heating, chemical processes, or desalination of sea water.The use of a HTR-type reactor in a co-generating way can value up to 90% of the energy produced. The high burn-up is due to the technology of HTR-type fuel that is based on encapsulation of fuel balls with heat-resisting materials. The nuclear fuel of Fort-Saint-Vrain unit (Usa) has reached values of burn-ups from 100.000 to 120.000 MWj/t. It is shown that the quantity of unloaded spent fuel can be divided by 4 for the same amount of electricity produced, in the case of the GT-MHR project in comparison with a light water reactor. (A.C.)

  7. State of development of high temperature gas-cooled reactors in foreign countries

    International Nuclear Information System (INIS)

    Sudo, Yukio

    1990-01-01

    Emphasis has been placed in the development of high temperature gas-cooled reactors on high thermal efficiency as power reactors and the reactor from which nuclear heat can be utilized. In U.K., as the international project 'Dragon Project', the experimental Dragon reactor for research use with 20 MWt output and exit coolant temperature 750 deg C was constructed, and operated till 1976. Coated fuel particles were developed. In West Germany, the experimental power reactor AVR with 46 MWt and 15 MWe output was operated till 1988. The prototype power reactor THTR-300 with 300 MWe output and 750 deg C exit temperature is in commercial operation. In USA, the experimental power reactor Peach Bottom reactor with 40 MWe output and 728 deg C exit temperature was operated till 1974. The prototype Fort Saint Vrain power reactor with 330 MWe output and 782 deg C exit temperature was operated till 1989. In USSR, the modular VGM with 200 MWh output is at the planning stage. Also in China, high temperature gas-cooled reactors are at the design stage. Switzerland has taken part in various international projects. (K.I.)

  8. Design of a new research reactor : 1st year conceptual design

    International Nuclear Information System (INIS)

    Park, Cheol; Lee, B. C.; Chae, H. T.

    2004-01-01

    A new research reactor model satisfying the strengthened regulatory environments and the changed circumstances around nuclear society should be prepared for the domestic and international demand of research reactor. This can also lead to the improvement of technologies and fostering manpower obtained during the construction and the operation of HANARO. In this aspect, this study has been launched and the 1st year conceptual design has been carried out in 2003. The major tasks performed at the first year of conceptual design stage are as follows; Establishments of general design requirements of research reactors and experimental facilities, Establishment of fuel and reactor core concepts, Preliminary analysis of reactor physics and thermal-hydraulics for conceptual core, Conceptual design of reactor structure and major systems, International cooperation to establish foundations for exporting

  9. Investigation on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin

    2013-10-22

    Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: • Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. • Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. • Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. • Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the

  10. The gas-cooled high temperature reactor: perspectives, problems and programmes

    International Nuclear Information System (INIS)

    Beckurts, K.H.; Engelmann, P.; Erb, D.E.

    1977-01-01

    For nearly 20 years, extensive research and development programs on Helium-cooled high-temperature reactors (HTR) have been carried out in several countries of the world, in particular in Germany and in the United States. This reactor system offers major potential advantages as a source of electricity or of nuclear process heat: it shows high nuclear fuel conversion efficiency, permitting a better utilization of uranium and in particular of thorium resources; it offers a high degree of inherent nuclear safety and thus a good potential for adoption to very strict safety standards; it permits high-efficiency electricity generation using either the indirect steam or the direct Helium cycle; dry air cooling can be employed without major economic penalties; it permits direct use of the nuclear heat for the production of gaseous or liquid secondary fuels from coal and other fossil fuels or - on a more extended time scale - by thermochemical water splitting. As a result of the longstanding efforts, satisfactory solutions have been found for many of the basic problems of this new reactor system, particularly in the field of high-temperature fuels and materials technology. Three small experimental plants - Peach Bottom in USA, Dragon in England, and AVR in Germany - have been operated successfully over extended periods of time. The AVR is still in operation; since 1974 it has performed satisfactorily with an average gas outlet temperature of 950 0 C. Prototype steam-cycle plants of 300 MW(e) are underway at Fort St. Vrain, USA (full-power operation scheduled for 1977), and at Schmehausen, Germany (scheduled for 1979). Major delays have occured in the construction and commissioning of these plants; they are due to various reasons and do not reveal specific problems of the HTR. Commercial market introduction of the steam-cycle electricity generating system has been attempted, but the first approach has not been successfull. Major effects by both government and industry are

  11. Nuclear Data Measurements for 21st Century Reactor Physics Applications

    Energy Technology Data Exchange (ETDEWEB)

    Rahmat Aryaeinejad; Jerald D. Cole; Mark W. Drigert; James K. Jewell; Christopher A. McGrath; David W. Nigg; Edward L. Reber

    2003-03-01

    The United States Department of Energy (DOE), Office of Nuclear Energy (NE) has embarked on a long-term program to significantly advance the science and technology of nuclear energy. This is in response to the overall national plan for accelerated development of domestic energy resources on several fronts, punctuated by recent dramatic events that have emphasized the need for the US to reduce its dependence on foreign petroleum supplies. Key aspects of the DOE-NE agenda are embodied in the Generation-IV (Gen-IV) advanced nuclear energy systems development program and in the Advanced Fuel Cycle (AFC) program. The planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current nuclear power reactor systems as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The success of the overall NE effort will depend not only on sophisticated system development and engineering, but also on the advances in the supporting sciences and technologies. Of these, one of the most important is the improvement of the relevant fundamental nuclear science data bases, especially the evaluated neutron interaction cross section files that serve as the foundation of all reactor system designs, operating strategies, and fuel cycle engineering activities. The new concepts for reactors and fuel cycles involve the use of transuranic nuclides that were previously of little interest, and where experimentally measured information is lacking. The current state of the cross section database for some of these nuclides is such that design computations for advanced fast-spectrum reactor systems and fuel cycles that incorporate such materials in significant quantities are meaningful only for approximate conceptual applications. No actual system could reliably be designed according to currently accepted standards, nor

  12. Nuclear Data Measurements for 21st Century Reactor Physics Applications

    International Nuclear Information System (INIS)

    Rahmat Aryaeinejad; Jerald D. Cole; Mark W. Drigert; James K. Jewell; Christopher A. McGrath; David W. Nigg; Edward L. Reber

    2003-01-01

    The United States Department of Energy (DOE), Office of Nuclear Energy (NE) has embarked on a long-term program to significantly advance the science and technology of nuclear energy. This is in response to the overall national plan for accelerated development of domestic energy resources on several fronts, punctuated by recent dramatic events that have emphasized the need for the US to reduce its dependence on foreign petroleum supplies. Key aspects of the DOE-NE agenda are embodied in the Generation-IV (Gen-IV) advanced nuclear energy systems development program and in the Advanced Fuel Cycle (AFC) program. The planned efforts involve near-term and intermediate-term improvements in fuel utilization and recycling in current nuclear power reactor systems as well as the longer-term development of new nuclear energy systems that offer much improved fuel utilization and proliferation resistance, along with continued advances in operational safety. The success of the overall NE effort will depend not only on sophisticated system development and engineering, but also on the advances in the supporting sciences and technologies. Of these, one of the most important is the improvement of the relevant fundamental nuclear science data bases, especially the evaluated neutron interaction cross section files that serve as the foundation of all reactor system designs, operating strategies, and fuel cycle engineering activities. The new concepts for reactors and fuel cycles involve the use of transuranic nuclides that were previously of little interest, and where experimentally measured information is lacking. The current state of the cross section database for some of these nuclides is such that design computations for advanced fast-spectrum reactor systems and fuel cycles that incorporate such materials in significant quantities are meaningful only for approximate conceptual applications. No actual system could reliably be designed according to currently accepted standards, nor

  13. Radiolytic reactions in the coolant of helium cooled reactors

    International Nuclear Information System (INIS)

    Tingey, G.L.; Morgan, W.C.

    1975-01-01

    The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

  14. 1st International School of Fusion Reactor Technology "Ettore Majorana"

    CERN Document Server

    Knoepfel, Heinz; Safety, Environmental Impact and Economic Prospects of Nuclear Fusion

    1990-01-01

    This book contains the lectures and the concluding discussion of the "Seminar on Safety, Environmental Impact, and Economic Prospects of Nuclear Fusion", which was held at Erice, August 6-12, 1989. In selecting the contributions to this 9th meeting held by the International School of Fusion Reactor Technology at the E. Majorana Center for Scientific Cul­ ture in Erice, we tried to provide a comprehensive coverage of the many interre­ lated and interdisciplinary aspects of what ultimately turns out to be the global acceptance criteria of our society with respect to controlled nuclear fusion. Consequently, this edited collection of the papers presented should provide an overview of these issues. We thus hope that this book, with its extensive subject index, will also be of interest and help to nonfusion specialists and, in general, to those who from curiosity or by assignment are required to be informed on these as­ pects of fusion energy.

  15. Nuclear Energy Center: upper St. Lawrence region. Part I. Siting. Part II. Fort Drum surrogate site, description and impact assessment. Part III. Dispersed sites impact assessment and comparison with the NEC

    Energy Technology Data Exchange (ETDEWEB)

    Merry, P.A.; Luner, C.; Hong, S.W.; Canham, H.O.; Boggs, J.F.; McCool, T.P.

    1976-12-01

    This report is one of many supporting documents used by the Nuclear Regulatory commission in the preparation of the Nuclear Energy Center Site Survey (NECSS) mandated by Congress. While the overall study focuses on the feasibility and practicability of nuclear energy centers (NECs), this report is directed towards choosing a suitable surrogate site in the upper St. Lawrence region of New York State, assessing the probable impacts associated with construction and operation of the NEC, and comparing these impacts with those associated with small dispersed nuclear power stations. The upper St. Lawrence region is surveyed to identify a specific site that might be suitable for a surrogate NEC. Several assumptions about the basic design of an NEC are delineated, and a general overview of the characteristics of the region is given. The Fort Drum Military Reservation is chosen as a suitable surrogate site. Fort Drum and the surrounding area are described in terms of land use and population patterns, terrestrial and aquatic ecology, water use and quality, meteorology, institutional framework, and socioeconomic structure. The impacts associated with NEC development are assessed. Then the impacts associated with smaller dispersed nuclear power stations located throughout New York State are assessed and compared with the impacts associated with the NEC. Finally, the impacts due to development of the transmission line networks associated with the NEC and with the dispersed power stations are assessed and compared.

  16. The Budapest research reactor as an advanced research facility for the early 21st century

    International Nuclear Information System (INIS)

    Vidovszky, I.

    2001-01-01

    The Budapest Research Reactor, Hungary's first nuclear facility was originally put into operation in 1959. The reactor serves for: basic and applied research, technological and commercial applications, education and training. The main goal of the reactor is to serve neutron research. This unique research possibility is used by a broad user community of Europe. Eight instruments for neutron scattering, radiography and activation analyses are already used, others (e.g. time of flight spectrometer, neutron reflectometer) are being installed. The majority of these instruments will get a much improved utilization when the cold neutron source is put into operation. In 1999 the Budapest Research Reactor was operated for 3129 full power hours in 14 periods. The normal operation period took 234 hours (starting Monday noon and finishing Thursday morning). The entire production for the year 1999 was 1302 MW days. This is a slightly reduced value, due to the installation of the cold neutron source. For the year 2000 a somewhat longer operation is foreseen (near to 4000 hours), as the cold neutron source will be operational. The operation of the reactor is foreseen at least up to the end of the first decade of the 21 st century. (author)

  17. A role of small reactors in the latter part of 21st century

    International Nuclear Information System (INIS)

    Sekimoto, H.

    2004-01-01

    In the year 2002 and 2003 the Japanese Ministry of Education, Culture, Sports, Science and Technology started the Priority Assistance for the Formation of Worldwide Renowned Centers of Research - The 21st Century Center of Excellence (COE) Program. A program proposed by Tokyo Institute of Technology Innovative Nuclear Energy Systems for Sustainable Development of the World was selected as only one program in nuclear engineering. In this program the system of nuclear energy park and small reactors will be intensively investigated. Here the small reactors are constructed in the nuclear energy park and transferred to the site and set there. The reactor vessel is sealed and can not be opened at the site. It is excellent from the nonproliferation point of view. It is also good from safety and easy maintenance, since refueling is not performed at the site. At the end of the reactor life it is replace by a new one. The old one will be shipped to the nuclear energy park. There is not radioactive wastes left at the site, the site is free from the waste problems. At the conference presented are the details of the considered nuclear energy park and a proposed design of the small reactor.(author)

  18. Moving into the 21st century - The United States' Research Reactor Spent Nuclear Fuel Acceptance Program

    International Nuclear Information System (INIS)

    Huizenga, David G.; Mustin, Tracy P.; Saris, Elizabeth C.; Reilly, Jill E.

    1999-01-01

    Since 1996, when the United States Department of Energy and the Department of State jointly adopted the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel, twelve shipments totaling 2,985 MTR and TRIGA spent nuclear fuel assemblies from research reactors around the world have been accepted into the United States. These shipments have contained approximately 1.7 metric tons of HEU and 0.6 metric tons of LEU. Foreign research reactor operators played a significant role in this success. A new milestone in the acceptance program occurred during the summer of 1999 with the arrival of TRIGA spent nuclear fuel from Europe through the Charleston Naval Weapons Station via the Savannah River Site to the Idaho National Engineering and Environmental Laboratory. This shipment consisted of five casks of TRIGA spent nuclear fuel from research reactors in Germany, Italy, Slovenia, and Romania. These casks were transported by truck approximately 2,400 miles across the United States (one cask packaged in an ISO container per truck). Drawing upon lessons learned in previous shipments, significant technical, legal, and political challenges were addressed to complete this cross-country shipment. Other program activities since the last RERTR meeting have included: formulation of a methodology to determine the quantity of spent nuclear fuel in a damaged condition that may be transported in a particular cask (containment analysis for transportation casks); publication of clarification of the fee policy; and continued planning for the outyears of the acceptance policy including review of reactors and eligible material quantities. The United States Foreign Research Reactor Spent Nuclear Fuel Acceptance Program continues to demonstrate success due to the continuing commitment between the United States and the research reactor community to make this program work. We strongly encourage all eligible research reactors to decide as soon as possible to

  19. The nuclear renaissance and AREVA's reactor designs for the 21st century. EPR and SWR-1000

    International Nuclear Information System (INIS)

    Stosic, Z.V.

    2007-01-01

    Hydro and nuclear energy are the most environmentally benign way of producing electricity on a large scale. Nuclear generated electricity releases 38 times fewer greenhouse gases than coal, 27 times fewer than oil and 15 times fewer than natural gas [9]. On a global scale nuclear power annually saves about 10% of the global CO 2 emission. European nuclear power plants save amount of CO 2 emissions corresponding with the annual emission of CO 2 from all European passenger cars [16]. Also, that is approximately twice the total estimated quantity to be avoided in Europe under the Kyoto Protocol during the period 2008-2012. In respect to main drivers - such as concerns of the global warming effect, population growth, and future energy supply shortfall, low operating costs, reduced dependence on imported gas - it is clear that 30 new nuclear reactors currently being constructed in 11 countries and another 35 and more planed during next 10 years confirm the nuclear renaissance. Participation in the construction of 100 reactors out of 443 worldwide operated in January 2006 and supplying fuel to 148 of them AREVA helps meet the 21 st century's greatest challenges: making energy available to all, protecting the planet, and acting responsibly towards future generations. With EPR and SWR-1000, AREVA NP has developed advanced design concepts of Generation III+ nuclear reactors which fully meet the most stringent requirements in terms of nuclear safety, operational reliability and economic performance. (author)

  20. Interim development report: engineering-scale HTGR fuel particle crusher

    International Nuclear Information System (INIS)

    Baer, J.W.; Strand, J.B.

    1978-09-01

    During the reprocessing of HTGR fuel, a double-roll crusher is used to fracture the silicon carbide coatings on the fuel particles. This report describes the development of the roll crusher used for crushing Fort-St.Vrain type fissile and fertile fuel particles, and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. Recommendations are made for design improvements and further testing

  1. Gas-cooled reactor technology safety and siting. Report of a technical committee meeting. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1990-07-01

    At the invitation of the Government of the Union of Soviet Socialist Republics, the Eleventh International Conference on the HTGR and the IAEA Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting were held in Dimitrovgrad, USSR, on June 21-23, 1989. The Technical Committee Meeting provided the Soviet delegates with an opportunity to display the breadth of their program on HTGRs to an international audience. Nearly one-half of the papers were presented by Soviet participants. Among the highlights of the meeting were the following: the diverse nature and large magnitude of the Soviet research and development program on high temperature gas-cooled reactors; the Government approval of the budget for the construction of the 30 MWt High Temperature Test Reactor (HTTR) in Japan (The schedule contemplates a start of construction in spring 1990 on a site at the Oarai Research Establishment and about a five year construction period.); disappointment in the announced plans to shutdown both the Fort St. Vrain (FSV) plant in the United States (US) and the Thorium High Temperature Reactor (THTR-300) in Germany (These two reactors presently represent the only operating HTGRs in the world since the AVR plant in Juelich, Germany, was also shutdown at the end of 1988.); the continuing negotiations between Germany and the USSR on the terms of the co-operation between the two countries for the construction of a HTR Module supplemented by joint research and development activities aimed at increasing coolant outlet temperatures from 750 deg. C to 950 deg. C; the continued enthusiasm displayed by both the US and German representatives for the potential of the small modular designs under development in both countries and the ability for these designs to meet the stringent requirements demanded for the future expansion of nuclear power; the combining of the HTGR technology interest of ABB-Atom and Siemens in Germany into a joint enterprise, HTR GmbH, in May 1989

  2. Gas-cooled reactor technology safety and siting. Report of a technical committee meeting. Working material

    International Nuclear Information System (INIS)

    1990-01-01

    At the invitation of the Government of the Union of Soviet Socialist Republics, the Eleventh International Conference on the HTGR and the IAEA Technical Committee Meeting on Gas-Cooled Reactor Technology, Safety and Siting were held in Dimitrovgrad, USSR, on June 21-23, 1989. The Technical Committee Meeting provided the Soviet delegates with an opportunity to display the breadth of their program on HTGRs to an international audience. Nearly one-half of the papers were presented by Soviet participants. Among the highlights of the meeting were the following: the diverse nature and large magnitude of the Soviet research and development program on high temperature gas-cooled reactors; the Government approval of the budget for the construction of the 30 MWt High Temperature Test Reactor (HTTR) in Japan (The schedule contemplates a start of construction in spring 1990 on a site at the Oarai Research Establishment and about a five year construction period.); disappointment in the announced plans to shutdown both the Fort St. Vrain (FSV) plant in the United States (US) and the Thorium High Temperature Reactor (THTR-300) in Germany (These two reactors presently represent the only operating HTGRs in the world since the AVR plant in Juelich, Germany, was also shutdown at the end of 1988.); the continuing negotiations between Germany and the USSR on the terms of the co-operation between the two countries for the construction of a HTR Module supplemented by joint research and development activities aimed at increasing coolant outlet temperatures from 750 deg. C to 950 deg. C; the continued enthusiasm displayed by both the US and German representatives for the potential of the small modular designs under development in both countries and the ability for these designs to meet the stringent requirements demanded for the future expansion of nuclear power; the combining of the HTGR technology interest of ABB-Atom and Siemens in Germany into a joint enterprise, HTR GmbH, in May 1989

  3. Fortællingen

    DEFF Research Database (Denmark)

    Hejlsted, Annemette

    Fortællingen - teori og analyse introducerer til teorier om fortællingen og præsenterer et sæt af analytiske tilgange til fortællinger af enhver art. Bogen lægger vægt på læsersynsvinklen og retter opmærksomheden mod de vilkår for menings- og betydningsdannelse, der kendetegner fortællingen. Begr....... Begreber om plot, fortællingens verden, karakterer, fortæller, modus og genre behandles, og deres anvendelse demonstreres på dansk og nordisk litteratur - med inddragelse af eksempler fra film og tv-reklamer....

  4. The WIMS-E module W-FORTE

    International Nuclear Information System (INIS)

    Roth, M.J.

    1983-09-01

    There are three distinct versions of the WIMS lattice cell program. WIMS-E is the most general, WIMSD4 is restricted to clusters or to one dimensional slab or annular geometry, and LWRWIMS is designed principally for light water reactor geometries. W-FORTE is used to transfer data from WIMSD4 or LWRWIMS to WIMS-E. A description of the W-FORTE module is given, and includes the relevant data for WIMSD4, LWRWIMS and W-FORTE. (UK)

  5. Supporting the national energy needs for the early 21st century with the advanced liquid metal reactor system (ALMRS)

    International Nuclear Information System (INIS)

    Hutchins, B.A.; Quinn, J.E.; Thompson, M.L.

    1991-01-01

    This paper presents a cost effective approach to providing a major contribution to the electricity needs of the United States in the early 21st century through an integrated Advanced Liquid Metal Reactor System (ALMRS). This system has several synergistic components which are under development by the United States Department of Energy (DOE): the modular, passively safe ALMR reactor design; metal fuel recycle (aka IFR); and the processing of LWR spent fuel to use as startup fuel for the ALMR. Each of these components contributes to an overall system behavior that will be able to provide an important portion of the United States' electrical energy needs beginning about the year 2010, while at the same time translating some fuel wastes of the LWR spent fuel to an asset. This paper describes each of these components and their synergism. Economic projections and busbar costs for this system are also presented

  6. Personnel radiation exposure in HTGR plants

    International Nuclear Information System (INIS)

    Su, S.; Engholm, B.A.

    1981-01-01

    Occupational radiation exposures in high-temperature gas-cooled reactor (HTGR) plants were assessed. The expected rate of dose accumulations for a large HTGR steam cycle unit is 0.07 man-rem/MW(e)y, while the design basis is 0.17 man-rem/MW(e)y. The comparable figure for actual light water reactor experience is 1.3 man-rem/MW(e)y. The favorable HTGR occupational exposure is supported by results from the Peach Bottom Unit No. 1 HTGR and Fort St. Vrain HTGR plants and by operating experience at British gas-cooled reactor stations

  7. Apollo - An advanced fuel fusion power reactor for the 21st century

    International Nuclear Information System (INIS)

    Kulcinski, G.L.; Emmert, G.A.; Blanchard, J.P.

    1989-01-01

    A preconceptual design of a tokamak reactor fueled by a D-He-3 plasma is presented. A low aspect ratio (A=2-4) device is studied here but high aspect ratio devices (A > 6) may also be quite attractive. The Apollo D-He-3 tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The overall efficiency ranges from 37 to 52% depending on whether the bremsstrahlung energy is utilized. The low neutron wall loading (0.1 MW/m/sup 2/) allows a permanent first wall to be designed and the low nuclear decay heat enables the reactor to be classed as inherently safe. The cost of electricity from Apollo is > 40% lower than electricity from a similar sized DT reactor

  8. Development and engineering plan for graphite spent fuels conditioning program

    International Nuclear Information System (INIS)

    Bendixsen, C.L.; Fillmore, D.L.; Kirkham, R.J.; Lord, D.L.; Phillips, M.B.; Pinto, A.P.; Staiger, M.D.

    1993-09-01

    Irradiated (or spent) graphite fuel stored at the Idaho Chemical Processing Plant (ICPP) includes Fort St. Vrain (FSV) reactor and Peach Bottom reactor spent fuels. Conditioning and disposal of spent graphite fuels presently includes three broad alternatives: (1) direct disposal with minimum fuel packaging or conditioning, (2) mechanical disassembly of spent fuel into high-level waste and low-level waste portions to minimize geologic repository requirements, and (3) waste-volume reduction via burning of bulk graphite and other spent fuel chemical processing of the spent fuel. A multi-year program for the engineering development and demonstration of conditioning processes is described. Program costs, schedules, and facility requirements are estimated

  9. A comparison of integral block and tubular interacting fuel element concepts for low enrichment HTR

    Energy Technology Data Exchange (ETDEWEB)

    Desoisa, J A

    1972-04-15

    The tubular interacting fuel element has to date been the favoured U.K. high temperature reactor design. Recent attempts to lower fuel costs and the progress of the Fort St. Vrain reactor has focussed attention on alternative designs, and in particular on the attractive design simplicity of the integral block concept. The aim of this investigation is to compare the merits of both concepts from fuel cycle cost and thermal performance viewpoints and to determine whether optimization of the integral block concept leads to changes in the current design values of (a) fuel density, (b) Nc/Nu, and/or (c) mean discharge irradiation within the framework of present design limits.

  10. Site-Specific, Covalent Immobilization of Dehalogenase ST2570 Catalyzed by Formylglycine-Generating Enzymes and Its Application in Batch and Semi-Continuous Flow Reactors

    Directory of Open Access Journals (Sweden)

    Hui Jian

    2016-07-01

    Full Text Available Formylglycine-generating enzymes can selectively recognize and oxidize cysteine residues within the sulfatase sub motif at the terminus of proteins to form aldehyde-bearing formylglycine (FGly residues, and are normally used in protein labeling. In this study, an aldehyde tag was introduced to proteins using formylglycine-generating enzymes encoded by a reconstructed set of the pET28a plasmid system for enzyme immobilization. The haloacid dehalogenase ST2570 from Sulfolobus tokodaii was used as a model enzyme. The C-terminal aldehyde-tagged ST2570 (ST2570CQ exhibited significant enzymological properties, such as new free aldehyde groups, a high level of protein expression and improved enzyme activity. SBA-15 has widely been used as an immobilization support for its large surface and excellent thermal and chemical stability. It was functionalized with amino groups by aminopropyltriethoxysilane. The C-terminal aldehyde-tagged ST2570 was immobilized to SBA-15 by covalent binding. The site-specific immobilization of ST2570 avoided the chemical denaturation that occurs in general covalent immobilization and resulted in better fastening compared to physical adsorption. The site-specific immobilized ST2570 showed 3-fold higher thermal stability, 1.2-fold higher catalytic ability and improved operational stability than free ST2570. The site-specific immobilized ST2570 retained 60% of its original activity after seven cycles of batch operation, and it was superior to the ST2570 immobilized to SBA-15 by physical adsorption, which loses 40% of its original activity when used for the second time. It is remarkable that the site-specific immobilized ST2570 still retained 100% of its original activity after 10 cycles of reuse in the semi-continuous flow reactor. Overall, these results provide support for the industrial-scale production and application of site-specific, covalently immobilized ST2570.

  11. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    International Nuclear Information System (INIS)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost

  12. Approach to the HTGR core outlet temperature measurements in the United States

    International Nuclear Information System (INIS)

    Franklin, R.; Rodriguez, C.

    1982-06-01

    The High Temperature Gas-Cooled Reactor (HTGR) constructed at Fort St. Vrain Colorado (330 MWe) used Geminol thermocouples to measure the primary coolant temperature at the core outlet. The primary coolant (helium) is heated by the graphite core to temperatures in the range of 700 deg. to 750 deg. C. The combination of the high temperature, high flow rate and radiation at the core outlet area makes it difficult to obtain accurate temperature measurements. The Geminol thermocouples installed in the Fort St. Vrain reactor have provided accurate data for several years of power operation without any failures. The indicated temperature of the core outlet thermocouples agrees with a ''traversing'' thermocouple measurement to within +-2 deg. C. The Geminol thermocouple wire was provided by the Driver-Harris Company and is similar to the chromel versus alumel thermocouple. Geminol wire is no longer distributed and on future designs, chromel versus alumel wire will be used. The next large HTGR design, which is being performed with funding support from the United States Department of Energy, will incorporate replaceable thermocouples. The thermocouples used in the Fort St. Vrain reactor were permanently installed and large in diameter (6.35 mm) to insure good reliability. The replaceable thermocouples to be used in the next large reactor will be smaller in diameter (3.18 mm). These replaceable thermocouples will be inserted into the core outlet area through long curved guide tubes that are permanently installed. These guide tubes are as long as 18 meters and must be curved to reach the core outlet regions. Tests were conducted to prove that the thermocouples could be inserted and removed through the long curved guide tubes. (author)

  13. Fortæller

    DEFF Research Database (Denmark)

    Larsen, Gorm

    2012-01-01

    Siden Gerard Genettes ”Discours du récit” (1972) er distinktionen mellem hvem, der taler, og hvem, der ser, blevet cementeret som et grundparadigme i narratologien og litteraturteorien. Genettes pointe var, at den etablerede narrative teori – som fx Wayne C. Booths The Rhetoric of Fiction (1961...... narratologi blevet forsøgt udfordret, enten fordi det hævdes, at en tekst ikke nødvendigvis er udstyret med en fortæller, eller fordi begrebet om fortæller antages at bero på en misvisende og reduktiv antropomorficering. Eller omvendt fordi der i Genettes begrebsdannelse ligger en forkastelse af...... forestillingen om en implicit forfatter (implied author) og dermed også en afvisning af en upålidelige fortæller. Kapitlet præsenterer begreberne fortæller og synsvinkel i narratologien med afsæt i Genettes bestemmelser og diskutere de problemer, der opstår i kølvandet herpå. Det være sig både de rent...

  14. Fortællerfiktionen

    DEFF Research Database (Denmark)

    Reitan, Rolf

    Bogen er en kritisk nærlæsning af Gérard Genettes Discours du récit og viser, hvorden den franske teoretiker løser og forenkler en række centrale problemer i traditionel fortælleteori, idet han uudtalt forudsætter et fiktionsbegreb, som han eksplicit afviser som narratologisk relevant. Det...

  15. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (Capsule 2) of Ulchin Nuclear Power Plant Unit 4

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2007-04-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin Unit 4 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.306E+18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.918 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.615E+18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 8.478E+18 and 1.673E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin Unit 4 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  16. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (CAPSULE 2) of Ulchin Nuclear Power Plant Unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2006-12-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.674E 18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.920 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.913E 18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 9.249E 18 and 1.834E 19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  17. 1st IAEA research coordination meeting on tritium retention in fusion reactor plasma facing components. October 5-6, 1995, Vienna, Austria. Summary report

    International Nuclear Information System (INIS)

    Langley, R.A.

    1995-12-01

    The proceedings and results of the 1st IAEA research Coordination Meeting on ''Tritium Retention in Fusion Reactor Plasma Facing Components'' held on October 5 and 6, 1995 at the IAEA Headquarters in Vienna are briefly described. This report includes a summary of presentations made by the meeting participants, the results of a data survey and needs assessment for the retention, release and removal of tritium from plasma facing components, a summary of data evaluation, and recommendations regarding future work. (author). 4 tabs

  18. HTGR plant availability and reliability evaluations. Volume I. Summary of evaluations

    International Nuclear Information System (INIS)

    Cadwallader, G.J.; Hannaman, G.W.; Jacobsen, F.K.; Stokely, R.J.

    1976-12-01

    The report (1) describes a reliability assessment methodology for systematically locating and correcting areas which may contribute to unavailability of new and uniquely designed components and systems, (2) illustrates the methodology by applying it to such components in a high-temperature gas-cooled reactor [Public Service Company of Colorado's Fort St. Vrain 330-MW(e) HTGR], and (3) compares the results of the assessment with actual experience. The methodology can be applied to any component or system; however, it is particularly valuable for assessments of components or systems which provide essential functions, or the failure or mishandling of which could result in relatively large economic losses

  19. Intervention in independent spent fuel storage facility license application proceedings for storage on the power plant site

    International Nuclear Information System (INIS)

    Jordan, J.

    1992-01-01

    This presentation summarizes the intervention in the Nuclear Regulatory Commission (NRC) licensing process for currently operating Independent Spent fuel Storage Installation (ISFSI) projects at Carolina Power and Light's Company's H.B. Robinson, Duke Power Company's Oconee, and Virginia Power Company's Surry. In addition, intervention at dry storage facilities that are currently under development are also described. The utilities and reactors include Baltimore Gas and Electric Company's Calvert Cliffs, Public Service Company of Colorado's Fort St. Vrain plant, Northern States Power Company's Prairie Island, Wisconsin Electric Power Company's Point Beach, and Consumers Power Company's Palisades

  20. HTGR plant availability and reliability evaluations. Volume I. Summary of evaluations

    Energy Technology Data Exchange (ETDEWEB)

    Cadwallader, G.J.; Hannaman, G.W.; Jacobsen, F.K.; Stokely, R.J.

    1976-12-01

    The report (1) describes a reliability assessment methodology for systematically locating and correcting areas which may contribute to unavailability of new and uniquely designed components and systems, (2) illustrates the methodology by applying it to such components in a high-temperature gas-cooled reactor (Public Service Company of Colorado's Fort St. Vrain 330-MW(e) HTGR), and (3) compares the results of the assessment with actual experience. The methodology can be applied to any component or system; however, it is particularly valuable for assessments of components or systems which provide essential functions, or the failure or mishandling of which could result in relatively large economic losses.

  1. FSV experience in support of the GT-MHR reactor physics, fuel performance, and graphite

    International Nuclear Information System (INIS)

    Baxter, A.M.; McEachern, D.; Hanson, D.L.; Vollman, R.E.

    1994-11-01

    The Fort St. Vrain (FSV) power plant was the most recent operating graphite-moderated, helium-cooled nuclear power plant in the United States. Many similarities exist between the FSV design and the current design of the GT-MHR. Both designs use graphite as the basic building blocks of the core, as structural material, in the reflectors, and as a neutron moderator. Both designs use hexagonal fuel elements containing cylindrical fuel rods with coated fuel particles. Helium is the coolant and the power densities vary by less than 5%. Since material and geometric properties of the GT-MHR core am very similar to the FSV core, it is logical to draw upon the FSV experience in support of the GT-MHR design. In the Physics area, testing at FSV during the first three cycles of operation has confirmed that the calculational models used for the core design were very successful in predicting the core nuclear performance from initial cold criticality through power operation and refueling. There was excellent agreement between predicted and measured initial core criticality and control rod positions during startup. Measured axial flux distributions were within 5% of the predicted value at the peak. The isothermal temperature coefficient at zero power was in agreement within 3%, and even the calculated temperature defect over the whole operating range for cycle 3 was within 8% of the measured defect. In the Fuel Performance area, fuel particle coating performance, and fission gas release predictions and an overall plateout analysis were performed for decommissioning purposes. A comparison between predicted and measured fission gas release histories of Kr-85m and Xe-138 and a similar comparison with specific circulator plateout data indicated good agreement between prediction and measured data. Only I-131 plateout data was overpredicted, while Cs-137 data was underpredicted

  2. 78 FR 32699 - Notice of Intent To Rule on Request to Release Airport Property at the Fort Worth Spinks Airport...

    Science.gov (United States)

    2013-05-31

    ... to Release Airport Property at the Fort Worth Spinks Airport, Fort Worth, Texas AGENCY: Federal Aviation Administration (FAA), DOT. ACTION: Notice of request to release airport property. SUMMARY: The FAA... the provisions of Section 125 of the Wendell H. Ford Aviation Investment Reform Act for the 21st...

  3. 1st IAEA research co-ordination meeting on 'plasma-material interaction data for mixed plasma facing materials in fusion reactors'. Summary report

    International Nuclear Information System (INIS)

    Janev, R.K.; Longhurst, G.

    1998-12-01

    The proceedings and conclusions of the 1st IAEA Research Co-ordination Meeting on 'Plasma-Material Interaction Data for Mixed Plasma Facing Materials in Fusion Reactors', held on December 19 and 20, 1998 at the IAEA Headquarters in Vienna, are briefly described. This report includes a summary of the presentations made by meeting participants, a review of the data availability and data needs in the areas from the scope of the Co-ordinated Research Project (CRP) on the subject of the meeting, and recommendations regarding the future work within this CRP. (author)

  4. 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.)

  5. List of reports on reactor safety research by BMFT, USNRC, EPRI and JSTA. Period under review: 1st January until 31st March 1978

    International Nuclear Information System (INIS)

    1978-05-01

    This list reviews reports from the Federal Republic of Germany, from the United States of America and from Japan concerning special problems in the field of Reactor Safety Research. According to the cooperation of the Bundesminister fuer Forschung und Technologie (BMFT) with the United States Regulatory Commisssion (USNRC), the Electric Power Research Institute (EPRI), and the Japan Science and Technology Agency (JSTA) these reports are available in the Gesellschaft fuer Reaktorsicherheit. The list pursues the following order: Country of origin, problem area concerned, according to the Reactor Safety Research Program of BMFT, reporting organisation. The list of reports appears quarterly. Requests for reports should be addressed to GRS, Forschungsbetreuung. Contractual view points have to be considered for the distribution of the reports. (orig.) [de

  6. 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.)

  7. Firmaets Største Bedrift

    DEFF Research Database (Denmark)

    Hansen, Peer Henrik

    fortæller, hvordan de tidligere modstandsfolk Arne Sejr og Niels Frommelt få år efter befrielsen etablerede deres såkaldte Firma for med støtte fra ledende danske politikere, Forsvarets Efterretningstjeneste og CIA at organisere en hemmelig krig mod Sovjetunionens danske støtter. Meget tyder på, at både...

  8. Fort Carson Wind Resource Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Robichaud, R.

    2012-10-01

    This report focuses on the wind resource assessment, the estimated energy production of wind turbines, and economic potential of a wind turbine project on a ridge in the southeastern portion of the Fort Carson Army base.

  9. The Design of High Reliability Magnetic Bearing Systems for Helium Cooled Reactor Machinery

    International Nuclear Information System (INIS)

    Swann, M.; Davies, N.; Jayawant, R.; Leung, R.; Shultz, R.; Gao, R.; Guo, Z.

    2014-01-01

    The requirements for magnetic bearing equipped machinery used in high temperature, helium cooled, graphite moderated reactor applications present a set of design considerations that are unlike most other applications of magnetic bearing technology in large industrial rotating equipment, for example as used in the oil and gas or other power generation applications. In particular, the bearings are typically immersed directly in the process gas in order to take advantage of the design simplicity that comes about from the elimination of ancillary lubrication and cooling systems for bearings and seals. Such duty means that the bearings will usually see high temperatures and pressures in service and will also typically be subject to graphite particulate and attendant radioactive contamination over time. In addition, unlike most industrial applications, seismic loading events become of paramount importance for the magnetic bearings system, both for actuators and controls. The auxiliary bearing design requirements, in particular, become especially demanding when one considers that the whole mechanical structure of the magnetic bearing system is located inside an inaccessible pressure vessel that should be rarely, if ever, disassembled over the service life of the power plant. Lastly, many machinery designs for gas cooled nuclear power plants utilize vertical orientation. This circumstance presents its own unique requirements for the machinery dynamics and bearing loads. Based on the authors’ experience with machine design and supply on several helium cooled reactor projects including Ft. St. Vrain (US), GT-MHR (Russia), PBMR (South Africa), GTHTR (Japan), and most recently HTR-PM (China), this paper addresses many of the design considerations for such machinery and how the application of magnetic bearings directly affects machinery reliability and availability, operability, and maintainability. Remote inspection and diagnostics are a key focus of this paper. (author)

  10. NSU Art Museum Fort Lauderdale | Art Museum in Fort Lauderdale

    Science.gov (United States)

    NSU Art Museum Fort Lauderdale Visit Admissions Hours & Admission Policies & Accessibility Airports Shop & Dine About the Café & Store Store Café Menu Art Exhibitions Currently on View Thursday 2-for-1 specials on wine and craft beer in the Museum Café, and hands-on art projects for all

  11. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  12. Criteria for the selection of graphites for HTR integral block fuel elements

    International Nuclear Information System (INIS)

    Knowles, A.N.

    1980-01-01

    This paper is concerned with the special requirements for integral block fuel elements of the type first used in the Fort St. Vrain reactor. The main idea of these elements is that the carrier block and separate graphite clad fuel pins are combined into a single monolith. This combination leads to lower fabrication costs and some improvement in the thermal performance (lower temperature difference between fuel and the surface of heat transfer into the coolant). The advent of block fuel for HTRs of the Fort St. Vrain type has placed a fresh emphasis on the selection of graphite for block manufacture in respect of physical properties. This is because the temperature distributions typical of such fuelled blocks lead to shutdown stresses close to the maximum the graphite can sustain without damage. Figures presented in this paper suggest that the physical properties of the graphite can play a relatively large part in reducing such stress levels and that guidance on the key requirements for suitable specifications is therefore particularly needed by the manufacturers of fuel block graphites. While graphites for fuel blocks have this special need for combinations of physical properties which lead to low thermal and shrinkage stresses, the other characteristics must also receive attention. A low graphite cost combined with good homogeneity in the brick, so that waste minimized, are still necessary, while isotropy is also very important

  13. Meeting of the Technical Working Group on Fast Reactors (TWG-FR) (41st Annual Meeting). Working Material

    International Nuclear Information System (INIS)

    2008-01-01

    The objectives of the meeting were to: - Exchange information on the national programmes on Fast Reactors (FR) and Accelerator Driven Systems (ADS); - Review the progress since the 40th TWG-FR Annual Meeting, including the status of the actions; - Consider meeting arrangements for 2008, 2009, 2010 and beyond; - Review the IAEA’s ongoing information exchange and coordinated research activities in the technical fields relevant to the TWG-FR (FRs and ADS), as well as coordination of the TWG-FR’s activities with other organizations and international initiatives; - Discuss future joint activities in view of IAEA’s Programme and Budget Cycles beyond 2008-2009

  14. Hot cell examination on the surveillance capsule of SA 533 cl. 1 reactor pressure vessel (1st test report)

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Yong Sun; Jung, Y. H.; Yoo, B. O.; Baik, S. J.; Oh, W. H.; Soong, W. S.; Hong, K. P

    2000-08-01

    The post-irradiated examinations such as impact test, tensile test, composition analysis and etc. were conducted to monitor and to evaluate the radiation-induced changes, so called radiation embrittlement, in the mechanical properties of ferritic materials. Those data should be applied to confirm safety as well as reliability of reactor pressure vessel. The scopes and contents of hot cell examination on the surveillance capsule are as follows; - Capsule transportation, cutting, dismantling and classification - Shim block and Dosimeter cutting and dismantling - Impact test - Tensile test - Composition analysis by EPMA - SEM observation on the fractured surface - Hardness test - Radwaste treatment.

  15. 1st quarterly report 1977

    International Nuclear Information System (INIS)

    1977-06-01

    The present report describes the activities carried out in the 1st quarter of 1977 at the Gesellschaft fuer Kernforschung in Karlsruhe or on its behalf in the framework of the fast breeder project (PSB). The problems and main results of the partial projects fuel rod development, materials testing, reactor physics, reactor safety and reactor technology are presented. (RW) [de

  16. 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.)

  17. Ecological Baseline, Fort Hood, Texas

    Science.gov (United States)

    1980-08-01

    cedar eTm (Uiimus crassifolia), Texas ash (Fraxinus texansis), and Texas persimmon ( Diospyros texana). Conversely, the two predominant tree species...Ilex decidua), Mex- ican buckeye (Ungnadia spjeciosa), and Texas persimmon ( Diospyros texana). Vines included greenbrier (Smilax bona-nox) and white...Hedgehey Cactus (Echinocereus sp.) has been observed on Fort Hood. Due to the brief period of flowering for this genus , the individual species were not

  18. Inventory of Forts in Indonesia

    Science.gov (United States)

    Rinandi, N.; Suryaningsih, F.

    2015-08-01

    The great archipelago in Indonesia with its wealthy and various nature, the products and commodities of tropic agriculture and the rich soil, was through the centuries a region of interest for other countries all over the world. For several reasons some of these countries came to Indonesia to establish their existence and tried to monopolize the trading. These countries such as the Portuguese, the Spanish, the Dutch and the British built strengthened trade stations which later became forts all over Indonesia to defend their interest. The archipelago of Indonesia possesses a great number of fortification-works as legacies of native rulers and those which were built by European trading companies and later became colonial powers in the 16th to the 19th centuries. These legacies include those specific structures built as a defence system during pre and within the period of World War II. These fortresses are nowadaysvaluable subjects, because they might be considered as shared heritage among these countries and Indonesia. It's important to develop a vision to preserve these particular subjects of heritage, because they are an interesting part of the Indonesian history and its cultural treasures. The Government of the Republic of Indonesia has national program to compile a comprehensive documentation of the existing condition of these various types of forts as cultural heritage. The result of the 3 years project was a comprehensive 442 forts database in Indonesia, which will be very valuable to the implementation of legal protection, preservation matters and adaptive re-use in the future.

  19. Operation Resettlement, Fort Chafee, Arkansas. 1st Psyop Bn

    Science.gov (United States)

    1981-01-01

    to the Civil Affairs Bat- talion Commander. As a special staff officer, the PSYOP Commander was required to attend the daily staff call (normally...fundener.tales que r.an .T, Primero. - l^a limpieza de las area it: vivienda y ti >ampai req cnsatilidad, nc tires papelec c basuras t...el terrcn envases de Iqahe, bandejas de carton, quc I’d. Vea e>: t I de basura o en las bolsas pl&sticas para qu< Mantenga su nuevc hogar

  20. HTGR nuclear heat source component design and experience

    International Nuclear Information System (INIS)

    Peinado, C.O.; Wunderlich, R.G.; Simon, W.A.

    1982-05-01

    The high-temperature gas-cooled reactor (HTGR) nuclear heat source components have been under design and development since the mid-1950's. Two power plants have been designed, constructed, and operated: the Peach Bottom Atomic Power Station and the Fort St. Vrain Nuclear Generating Station. Recently, development has focused on the primary system components for a 2240-MW(t) steam cycle HTGR capable of generating about 900 MW(e) electric power or alternately producing high-grade steam and cogenerating electric power. These components include the steam generators, core auxiliary heat exchangers, primary and auxiliary circulators, reactor internals, and thermal barrier system. A discussion of the design and operating experience of these components is included

  1. Sustainability Analysis of the Water Resources and Supply of the Vieux Fort Region of Saint Lucia

    Science.gov (United States)

    Coles, D.; Johnson, B.; Morgan, F.

    2005-05-01

    In the Vieux Fort region of the Caribbean island of St. Lucia, water needs are becoming acute. The water supply shortfalls during the dry season will continue to grow as population and development increase, unless action is taken. Actions to address the problem should include measures to optimize the present water delivery system and the development of a new supply, through new intakes, groundwater, or reservoir construction. An investigation into the potential for groundwater resources using electrical resistivity soundings indicated a likely pervasive, shallow aquitard of clay materials below the water table; the shallowness of this aquitard virtually precludes the existence of productive perched aquifers. Consequently, a model of Grande Riviere du Vieux Fort (Big Vieux Fort River) seasonal surface-water flow was developed, based on a digital elevation model and rainfall data, allowing us to analyze the possible productivity of any new intakes placed along the river. A specific site downstream of the present intake was recommended for potential development. Recommendations were given for short, medium and long-term development of the resources and supply of the Vieux Fort region of southern St. Lucia.

  2. 75 FR 55401 - Notice of Intent To Rule on Request To Release Airport Property at the Dallas/Fort Worth...

    Science.gov (United States)

    2010-09-10

    ... To Release Airport Property at the Dallas/Fort Worth International Airport, DFW Airport, TX AGENCY... airport property. SUMMARY: The FAA proposes to rule and invite public comment on the request for permanent... H. Ford Aviation Investment Reform Act for the 21st Century (AIR 21). DATES: Comments must be...

  3. 78 FR 9105 - Notice of Intent To Rule on Request To Release Airport Property at the Dallas/Fort Worth...

    Science.gov (United States)

    2013-02-07

    ... To Release Airport Property at the Dallas/Fort Worth International Airport, DFW Airport, TX AGENCY... Airport Property. SUMMARY: The FAA proposes to rule and invite public comment on the request for permanent... H. Ford Aviation Investment Reform Act for the 21st Century (AIR 21). DATES: Comments must be...

  4. Fort Davis National Historic Site : acoustical monitoring

    Science.gov (United States)

    2013-06-01

    During the summer of 2010 (September - October 2010), the Volpe Center collected baseline acoustical data at Fort Davis National Historic Site (FODA)at two sites deployed for approximately 30 days each. The baseline data collected during this period ...

  5. Renewable Energy Opportunities at Fort Hood, Texas

    Energy Technology Data Exchange (ETDEWEB)

    Solana, Amy E.; Warwick, William M.; Orrell, Alice C.; Russo, Bryan J.; Parker, Kyle R.; Weimar, Mark R.; Horner, Jacob A.; Manning, Anathea

    2011-11-14

    This report presents the results of Pacific Northwest National Laboratory's (PNNL) follow-on renewable energy (RE) assessment of Fort Hood. Fort Hood receives many solicitations from renewable energy vendors who are interested in doing projects on site. Based on specific requests from Fort Hood staff so they can better understand these proposals, and the results of PNNL's 2008 RE assessment of Fort Hood, the following resources were examined in this assessment: (1) Municipal solid waste (MSW) for waste-to-energy (WTE); (2) Wind; (3) Landfill gas; (4) Solar photovoltaics (PV); and (5) Shale gas. This report also examines the regulatory issues, development options, and environmental impacts for the promising RE resources, and includes a review of the RE market in Texas.

  6. Hydrologic Analysis of Fort Leonard Wood, Missouri

    Science.gov (United States)

    2015-08-01

    drainage areas are different, hydrological analysis will be conducted on the two basins individually. The results of the two analyses will be combined to...ER D C TR -1 5- 4 Environmental Quality and Installations Hydrologic Analysis of Fort Leonard Wood, Missouri En gi ne er R es ea rc h...Environmental Quality and Installations ERDC TR-15-4 August 2015 Hydrologic Analysis of Fort Leonard Wood, Missouri Michael L. Follum, Darla C. McVan

  7. Bent's Old Fort: Amphibians and Reptiles

    Science.gov (United States)

    Muths, E.

    2008-01-01

    Bent's Old Fort National Historic Site sits along the Arkansas River in the semi-desert prairie of southeastern Colorado. The USGS provided assistance in designing surveys to assess the variety of herpetofauna (amphibians and reptiles) resident at this site. This brochure is the results of those efforts and provides visitors with information on what frogs, toads, snakes and salamanders might be seen and heard at Bent's Old Fort.

  8. EPR: The reactor generation for the 21st century. SFEN-KTG congress 'The EPR Project'', Strasbourg, 13-14 Nov. 1994

    International Nuclear Information System (INIS)

    Anon.

    1995-01-01

    The foundation of the subsidiary in 1989, Nuclear Power International (NPI), brought together as partners the two formerly competing companies Siemens/KWU and Framatome who now are planning the novel reactor type within the framework of the joint EPR project. The project goal is to develop a novel French-German PWR reactor type for power generation, named European Pressurized Water Reactor (EPR). The article summarizes the main results of papers and discussions of the congress. (orig./UA) [de

  9. Review of tritium behavior in HTGR systems

    International Nuclear Information System (INIS)

    Gainey, B.W.

    1976-01-01

    The available experimental evidence from laboratory and reactor studies pertaining to tritium production, capture, release, and transport within an HTGR leading to release to the environment is reviewed. Possible mechanisms for release, capture, and transport are considered and a simple model was used to calculate the expected tritium release from HTGRs. Comparison with Federal regulations governing tritium release confirm that expected HTGR releases will be well within the allowable release limits. Releases from HTGRs are expected to be somewhat less than from LWRs based on the published LWR operating data. Areas of research deserving further study are defined but it is concluded that a tritium surveillance at Fort St. Vrain is the most immediate need

  10. Interim design report: fuel particle crushing

    International Nuclear Information System (INIS)

    Baer, J.W.; Strand, J.B.; Cook, E.J.; Miller, C.M.

    1977-11-01

    The double-roll fuel particle crusher was developed to fracture the silicon carbide coatings of Fort St. Vrain (FSV) fertile and fissile and large high-temperature gas-cooled reactor (LHTGR) fissile fuel particles. The report details the design task for the fuel particle crusher, including historical test information on double-roll crushers for carbide-coated fuels and the design approach selected for the cold pilot plant crusher, and shows how the design addresses the equipment goals and operational objectives. Design calculations and considerations are included to support the selection of crusher drive and gearing, the materials chosen for crushing rolls and housing, and the bearing selection. The results of the initial testing for compliance with design objectives and operational capabilities are also presented. 8 figures, 4 tables

  11. Licensing of HTGRs in the United States

    International Nuclear Information System (INIS)

    Fisher, C.R.; Orvis, D.D.

    1981-01-01

    The licensing history of the high-temperature gas-cooled reactor (HTGR) in the United States is given historical perspective. The experience began with the licensing of the Peach Bottom Atomic Power Station and extends to the continuing experience at the Fort St. Vrain Nuclear Generating Station. Additional experience was obtained from the licensing reviews in the mid-1970s of the large HTGR plants that were to be built by Philadelphia Electric Company and Delmarva Power and Light. Also, information was provided by the licensing review of the General Atomic standard plant by the U.S. Nuclear Regulatory Commission (NRC) at about the same time. These experiences are summarized in terms of the principal design criteria that were required by the regulatory authority for each project. These criteria include specification of the design basis accidents that were postulated for the plant safety analysis. Several technical issues raised by the NRC during their review of the large HTGR are presented. (author)

  12. Licensing of HTGRs in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, C. R.; Orvis, D. D. [General Atomic Co., San Diego, CA (USA)

    1981-01-15

    The licensing history of the high-temperature gas-cooled reactor (HTGR) in the United States is given historical perspective. The experience began with the licensing of the Peach Bottom Atomic Power Station and extends to the continuing experience at the Fort St. Vrain Nuclear Generating Station. Additional experience was obtained from the licensing reviews in the mid-1970s of the large HTGR plants that were to be built by Philadelphia Electric Company and Delmarva Power and Light. Also, information was provided by the licensing review of the General Atomic standard plant by the U.S. Nuclear Regulatory Commission (NRC) at about the same time. These experiences are summarized in terms of the principal design criteria that were required by the regulatory authority for each project. These criteria include specification of the design basis accidents that were postulated for the plant safety analysis. Several technical issues raised by the NRC during their review of the large HTGR are presented.

  13. Destruction of nuclear graphite using closed chamber incineration

    International Nuclear Information System (INIS)

    Senor, D.J.; Hollenberg, G.W.; Morgan, W.C.; Marianowski, L.G.

    1994-01-01

    Closed chamber incineration (CCI) is a novel technique by which irradiated nuclear graphite may be destroyed without the risk of radioactive cation release into the environment. The process utilizes an enclosed combustion chamber coupled with molten carbonate fuel cells (MCFCs). The transport of cations is intrinsically suppressed by the MCFCs, such that only the combustion gases are conducted through for release to the environment. An example CCI design was developed which had as its goal the destruction of graphite fuel elements from the Fort St. Vrain reactor (FSVR). By employing CCI, the volume of high level waste from the FSVR will be reduced by approximately 87 percent. Additionally, the incineration process will convert the SiC coating on the FSVR fuel particles to SiO 2 , thus creating a form potentially suitable for direct incorporation in a vitrification process stream. The design is compact, efficient, and makes use of currently available technology

  14. HTGR fuel element structural design consideration

    International Nuclear Information System (INIS)

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1987-01-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabilistic stress analysis techniques coupled with probabilistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistant with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the U.S.A. is discussed in the context of stress analysis uncertainty and structural criteria development. (author)

  15. HTGR fuel element structural design considerations

    International Nuclear Information System (INIS)

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1986-09-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabalistic stress analysis techniques coupled with probabalistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistent with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the USA is discussed in the context of stress analysis uncertainty and structural criteria development

  16. Comparison of activation cross section measurements and experimental techniques for fusion reactor technology. Summary report of the IAEA specialists' meeting held in St. Petersburg, Russia, 7 to 9 September 1994

    International Nuclear Information System (INIS)

    Pashchenko, A.B.

    1995-02-01

    The report contains the Summary of the IAEA Specialists' Meeting on ''Comparison of Activation Cross Section Measurements and Experimental Techniques for Fusion Reactor Technology''. The meeting was organized by the IAEA Nuclear Data Section (NDS) with co-operation and assistance of local organizers from the V.G. Khlopin Radium Institute, KRI, and held in St. Petersburg, Russia, from 7 to 9 September 1994. The aim of the meeting was to discuss and evaluate the preliminary results of the researches carried out in the framework of the international programme on Comparison of Activation Cross Section Measurements and Experimental Techniques for Fusion Reactor Technology coordinated by the IAEA Nuclear Data Section and to identify further measurements and actions of participating laboratories. The detailed conclusions and recommendations of the meeting are presented in Attachment 1 of the summary report. It was confirmed that for further comparison of experimental techniques the experimental groups at JAERI (Tokai, Japan), KRI (St. Petersburg, Russia), IPPE (Obninsk, Russia) and IEP (Debrecen, Hungary) will join in a collaborative program on comparing their measurement techniques and do measurements for reactions where discrepancies between their previous measurements exist. In cases where the JAERI results are the only existing data or deviate strongly from previous measurements, collaboration between KRI, IEP, IPPE and other institutions can consider measurements of these cross sections in order to clarify the situation. (author)

  17. FORTE spacecraft vibration mitigation. Final report

    International Nuclear Information System (INIS)

    Maly, J.R.

    1996-02-01

    This report documents work that was performed by CSA Engineering, Inc., for Los Alamos National Laboratory (LANL), to reduce vibrations of the FORTE spacecraft by retrofitting damped structural components into the spacecraft structure. The technical objective of the work was reduction of response at the location of payload components when the structure is subjected to the dynamic loading associated with launch and proto-qualification testing. FORTE is a small satellite that will be placed in orbit in 1996. The structure weighs approximately 425 lb, and is roughly 80 inches high and 40 inches in diameter. It was developed and built by LANL in conjunction with Sandia National Laboratories Albuquerque for the United States Department of Energy. The FORTE primary structure was fabricated primarily with graphite epoxy, using aluminum honeycomb core material for equipment decks and solar panel substrates. Equipment decks were bonded and bolted through aluminum mounting blocks to adjoining structure

  18. Renewable Energy Opportunities at Fort Polk, Louisiana

    Energy Technology Data Exchange (ETDEWEB)

    Solana, Amy E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Boyd, Brian K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Horner, Jacob A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Gorrissen, Willy J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Orrell, Alice C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Weimar, Mark R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hand, James R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Russo, Bryan J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Williamson, Jennifer L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2010-11-17

    This document provides an overview of renewable resource potential at Fort Polk, based primarily upon analysis of secondary data sources supplemented with limited on-site evaluations. This effort focuses on grid-connected generation of electricity from renewable energy sources and also on ground source heat pumps for heating and cooling buildings. The effort was funded by the U.S. Army Installation Management Command (IMCOM) as follow-on to the 2005 Department of Defense (DoD) Renewables Assessment. The site visit to Fort Polk took place on February 16, 2010.

  19. Renewable Energy Opportunities at Fort Sill, Oklahoma

    Energy Technology Data Exchange (ETDEWEB)

    Boyd, Brian K.; Hand, James R.; Horner, Jacob A.; Orrell, Alice C.; Russo, Bryan J.; Weimar, Mark R.; Nesse, Ronald J.

    2011-03-31

    This document provides an overview of renewable resource potential at Fort Sill, based primarily upon analysis of secondary data sources supplemented with limited on-site evaluations. This effort focuses on grid-connected generation of electricity from renewable energy sources and on ground source heat pumps for heating and cooling buildings. The effort was funded by the U.S. Army Installation Management Command (IMCOM) as follow-on to the 2005 Department of Defense (DoD) Renewables Assessment. The site visit to Fort Sill took place on June 10, 2010.

  20. Redmedial Action Plan for the Risk-Based Remediation of Site ST14 (SWMU 68), LPSTID 104819; the Former Base Refueling Area (A0C7); the French Underdrain System (SWMU 64); and the North Oil/Water Separator (SWMU 67), Carswell Air Force Base, Naval Air Station Fort Worth Joint Reserve Base, Texas. Volume 1: Report

    National Research Council Canada - National Science Library

    1997-01-01

    ...) to prepare a remedial action plan (RAP) in support of a risk-based remediation decision for soil and groundwater contaminated with fuel hydrocarbons at Site ST14 at Carswell Air Force Base (AFB), Texas...

  1. Defense.gov Special Report: Fort Hood Shooting

    Science.gov (United States)

    identify possible insider threats, Army Secretary John M. McHugh told lawmakers. Story Obama: Soldiers ," Army Secretary John M. McHugh told lawmakers. Story President Praises Swift Response to Fort Hood Remarks on Fort Hood Shooting at White House McHugh, Odierno Address Fort Hood Shooting Before Congress

  2. Smithsonian Marine Station (SMS) at Fort Pierce

    Science.gov (United States)

    share current Smithsonian research on the plants and animals of the Indian River Lagoon and marine Smithsonian Marine Station at Fort Pierce Website Search Box Search Field: SMS Website Search Twitter SMS Home › Welcome to the Smithsonian Marine Station Homepage slideshow Who We Are... The

  3. FORT NAMUTONI: FROM MILITARY STRONGHOLD TO TOURIST ...

    African Journals Online (AJOL)

    STRONGHOLD TO TOURIST CAMP. Col Dr Jan Ploeger*. "... this fortress was not just a white elephant, it was actually occupied and played a major role in the settlement of Germans in the far North." (own translation) - D.W. Krynauw Die Verhaal van. Namutoni, p 3. Introduction. Fort Namutoni, the last White outpost east of ...

  4. Den tabte fortælling

    DEFF Research Database (Denmark)

    Jørgensen, Kenneth Mølbjerg

    2008-01-01

    Ledelse er et af nøgleordene i fornyelsen af den offentlige sektor. Vi har imidlertid glemt et væsentligt aspekt af ledelse. Dette skyldes ikke mindst, at omgangsformen i dag er reguleret af information, mens den tidligere var reguleret af fortælleevnen. Evnen til dialog, indlevelse og nærvær er...

  5. Final safety analysis report for the irradiated fuels storage facility

    International Nuclear Information System (INIS)

    Bingham, G.E.; Evans, T.K.

    1976-01-01

    A fuel storage facility has been constructed at the Idaho Chemical Processing Plant to provide safe storage for spent fuel from two commercial HTGR's, Fort St. Vrain and Peach Bottom, and from the Rover nuclear rocket program. The new facility was built as an addition to the existing fuel storage basin building to make maximum use of existing facilities and equipment. The completed facility provides dry storage for one core of Peach Bottom fuel (804 elements), 1 1 / 2 cores of Fort St. Vrain fuel (2200 elements), and the irradiated fuel from the 20 reactors in the Rover program. The facility is designed to permit future expansion at a minimum cost should additional storage space for graphite-type fuels be required. A thorough study of the potential hazards associated with the Irradiated Fuels Storage Facility has been completed, indicating that the facility is capable of withstanding all credible combinations of internal accidents and pertinent natural forces, including design basis natural phenomena of a 10,000 year flood, a 175-mph tornado, or an earthquake having a bedrock acceleration of 0.33 g and an amplification factor of 1.3, without a loss of integrity or a significant release of radioactive materials. The design basis accident (DBA) postulated for the facility is a complete loss of cooling air, even though the occurrence of this situation is extremely remote, considering the availability of backup and spare fans and emergency power. The occurrence of the DBA presents neither a radiation nor an activity release hazard. A loss of coolant has no effect upon the fuel or the facility other than resulting in a gradual and constant temperature increase of the stored fuel. The temperature increase is gradual enough that ample time (28 hours minimum) is available for corrective action before an arbitrarily imposed maximum fuel centerline temperature of 1100 0 F is reached

  6. Final report for the 1st ex-vessel neutron dosimetry installations and evaluations for Kori unit 2 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-11-15

    This report describes a neutron fluence assessment performed for the Kori unit 2 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 20 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 2 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 20.

  7. Final report for the 1st ex-vessel neutron dosimetry installation and evaluations for Kori unit 4 reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Yoo, Choon Sung; Lee, Sam Lai; Chang, Kee Ok; Gong, Un Sik; Choi, Kwon Jae; Chang, Jong Hwa; Lim, Nam Jin; Hong, Joon Wha; Cheon, Byeong Jin

    2006-11-15

    This report describes a neutron fluence assessment performed for the Kori unit 4 pressure vessel belt line region based on the guidance specified in regulatory guide 1.190. In this assessment, maximum fast neutron exposures expressed in terms of fast neutron fluence (E>1 MeV) and iron atom displacements (dpa) were established for the belt line region of the pressure vessel. During cycle 16 of reactor operation, an ex-vessel neutron dosimetry program was instituted at Kori unit 4 to provide continuous monitoring of the belt line region of the reactor vessel. The use of the ex-vessel neutron dosimetry program coupled with available surveillance capsule measurements provides a plant specific data base that enables the evaluation of the vessel exposure and the uncertainty associated with that exposure over the service life of the unit. Ex-vessel neutron dosimetry has been evaluated at the conclusion of cycle 16.

  8. Fortælleværksteder

    DEFF Research Database (Denmark)

    Krøjer, Jo; Hutters, Camilla

    2009-01-01

    Unges valg af videregående uddannelse er omgærdet af forventninger. Forventninger til hvad man skal vælge. Forventninger til hvor lang tid, man skal være om at tage en uddannelse. Og forventninger til, hvad uddannelsen skal føre til. Artiklen præsenterer fortælleværkstedet, en metode til kollekti...... refleksioner over egne og adres forventninger til og tanker om uddannelsesvalg....

  9. Rail Outloading Capability Study, Fort Polk, Louisiana,

    Science.gov (United States)

    1977-06-01

    regardless of experience, to avoid wasted man -hours. The main problem at Fort Polk is that no blocking and bracing material stockpile exists and no...ti1 hottul only thtrough the 0111crinost hole; to defect within 20 days after it is determined to -tuit Owt ttrtk in tuse. III thle caste of classes 3...wheels, slipping, or similar trak (meh causes. 1 -------------------- (12) " Shelly spots" means a condition 2 ------------------------ % where a thin

  10. Renewable Energy Opportunities at Fort Hood, Texas

    Energy Technology Data Exchange (ETDEWEB)

    Chvala, William D.; Warwick, William M.; Dixon, Douglas R.; Solana, Amy E.; Weimar, Mark R.; States, Jennifer C.; Reilly, Raymond W.

    2008-06-30

    The document provides an overview of renewable resource potential at Fort Hood based primarily upon analysis of secondary data sources supplemented with limited on-site evaluations. The effort was funded by the U.S. Army Installation Management Command (IMCOM) as follow-on to the 2005 DoD Renewables Assessment. This effort focuses on grid-connected generation of electricity from renewable energy sources and also ground source heat pumps for heating and cooling buildings, as directed by IMCOM.

  11. FIND: Fort Calhoun Station, Unit 2

    International Nuclear Information System (INIS)

    Williams, W.H.

    1976-07-01

    This index is presented for the microfiche material of Docket 50548 which concerns the application of Omaha Public Power District to build and operate Fort Calhoun Station, Unit 2. The information includes both application and review material dated from September 1975 through March 1976. There are five amendments to the PSAR and one supplement to the ER which have been incorporated by reference into the respective reports. Docket RESAR-3 is used as a reference for portions of the PSAR

  12. Feltarbejde i Thule. Sammenfiltringen af steder, folk og fortællinger

    Directory of Open Access Journals (Sweden)

    Kirsten Hastrup

    2016-07-01

    Full Text Available På baggrund af lang tids arbejde i Thuleregionen i det nordvestligste Grønland vil jeg diskutere, hvordan steder, folk og fortællinger gensidigt former hinanden. ’Felten’ er således formateret af mange forhold, historiske og nutidige, naturlige og kulturelle, og man må besinde sig på feltens flydende form, selv når den ser mest solid ud. Steder er i sig selv flygtige; de opstår i mødet med mennesker, som tillægger dem betydning. Folk kan se nok så traditionelle ud, men de lever i samme verden som antropologen, der kommer for at lære af dem. Endelig er fortællingerne ikke stivnede vidnesbyrd om tidligere tider; de er tværtimod et vigtigt redskab i håndteringen af højst nutidige udfordringer, som kommer til syne i det endnu ufortalte. Bag fortællingen om Thule ligger en større diskussion af enhver felts plasticitet.

  13. 78 FR 17087 - Special Local Regulation; New River Raft Race, New River; Fort Lauderdale, FL

    Science.gov (United States)

    2013-03-20

    ...-AA08 Special Local Regulation; New River Raft Race, New River; Fort Lauderdale, FL AGENCY: Coast Guard... on the New River in Fort Lauderdale, Florida during the Rotary Club of Fort Lauderdale New River Raft... States during the Rotary Club of Fort Lauderdale New River Raft Race. On March 23, 2013, Fort Lauderdale...

  14. ACRR fission product release tests: ST-1 and ST-2

    International Nuclear Information System (INIS)

    Allen, M.D.; Stockman, H.W.; Reil, K.O.; Grimley, A.J.; Camp, W.J.

    1988-01-01

    Two experiments (ST-1 and ST-2) have been performed in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNLA) to obtain time-resolved data on the release of fission products from irradiated fuels under light water reactor (LWR) severe accident conditions. Both experiments were conducted in a highly reducing environment at maximum fuel temperatures of greater than 2400 K. These experiments were designed specifically to investigate the effect of increased total pressure on fission product release; ST-1 was performed at approximately 0.16 MPa and ST-2 was run at 1.9 MPa, whereas other parameters were matched as closely as possible. Release rate data were measured for Cs, I, Ba, Sr, Eu, Te, and U. The release rates were higher than predicted by existing codes for Ba, Sr, Eu, and U. Te release was very low, but Te did not appear to be sequestered by the zircaloy cladding; it was evenly distributed in the fuel. In addition, in posttest analysis a unique fuel morphology (fuel swelling) was observed which may have enhanced fission product release, especially in the high pressure test (ST-2). These data are compared with analytical results from the CORSOR correlation and the VICTORIA computer model

  15. Initiating Event Rates at U.S. Nuclear Power Plants. 1988 - 2013

    International Nuclear Information System (INIS)

    Schroeder, John A.; Bower, Gordon R.

    2014-01-01

    Analyzing initiating event rates is important because it indicates performance among plants and also provides inputs to several U.S. Nuclear Regulatory Commission (NRC) risk-informed regulatory activities. This report presents an analysis of initiating event frequencies at U.S. commercial nuclear power plants since each plant's low-power license date. The evaluation is based on the operating experience from fiscal year 1988 through 2013 as reported in licensee event reports. Engineers with nuclear power plant experience staff reviewed each event report since the last update to this report for the presence of valid scrams or reactor trips at power. To be included in the study, an event had to meet all of the following criteria: includes an unplanned reactor trip (not a scheduled reactor trip on the daily operations schedule), sequence of events starts when reactor is critical and at or above the point of adding heat, occurs at a U.S. commercial nuclear power plant (excluding Fort St. Vrain and LaCrosse), and is reported by a licensee event report. This report displays occurrence rates (baseline frequencies) for the categories of initiating events that contribute to the NRC's Industry Trends Program. Sixteen initiating event groupings are trended and displayed. Initiators are plotted separately for initiating events with different occurrence rates for boiling water reactors and pressurized water reactors. p-values are given for the possible presence of a trend over the most recent 10 years.

  16. Plant maintenance and advanced reactors, 2006

    Energy Technology Data Exchange (ETDEWEB)

    Agnihotri, Newal (ed.)

    2006-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Advanced plants to meet rising expectations, by John Cleveland, International Atomic Energy Agency, Vienna; A flexible and economic small reactor, by Mario D. Carelli and Bojan Petrovic, Westinghouse Electric Company; A simple and passively safe reactor, by Yury N. Kuznetsov, Research and Development Institute of Power Engineering (NIKIET), Russia; Gas-cooled reactors, by Jeffrey S. Merrifield, U.S. Nuclear Regulatory Commission; ISI project managment in the PRC, by Chen Chanbing, RINPO, China; and, Fort Calhoun refurbishment, by Sudesh Cambhir, Omaha Public Power District.

  17. Fort Mason Center: Pier 2 Project

    Energy Technology Data Exchange (ETDEWEB)

    Nester, Patrick [Fort Mason Center, San Francisco, CA (United States)

    2014-08-30

    The rooftop Photovoltaic (PV) panels and radiant piping project was constructed by Fort Mason Center as part of its $21 million comprehensive rehabilitation of the Pier 2 shed which include the shed’s electrical, natural gas and water systems. Fort Mason Center improved performance while reducing energy and water usage and costs to demonstrate the efficiencies and opportunities available to large multi-function facilities. The scalable demand of these facilities required a layered approach to conservation, control and production. The project employed a comprehensive retrofit of electrical natural gas, and plumbing systems to maximize efficiency and lower carbon footprint specifically to demonstrate the effectiveness of these strategies in a public setting with varied and diverse use. The project was completed in July 2014 and met the expected outcomes regarding increased comfort and operational efficiency throughout the Pier 2 shed as well as on site electrical generation of current consumption. The entire Pier 2 shed project won a 2015 California Preservation Foundation design award for historic rehabilitation.

  18. The hill forts and castle mounds in Lithuania: interaction between geodiversity and human-shaped landscape

    Science.gov (United States)

    Skridlaite, Grazina; Guobyte, Rimante; Satkunas, Jonas

    2015-04-01

    Lithuania is famous for its abundant, picturesque hill forts and castle mounds of natural origin. In Lithuania as well as in whole Europe the fortified hills were used as the society dwelling place since the beginning of the Late Bronze Age. Their importance increased when Livonian and Teutonic Orders directed a series of military campaigns against Lithuania with the aim of expansion of Christianity in the region at the end of 1st millennium AD, and they were intensively used till the beginning of the 15th c. when most of them were burned down during fights with the Orders or just abandoned due to the changing political and economical situation. What types of the geodiversity were used for fortified dwellings? The choice in a particular area depended on a variety of geomorphology left behind the retreating ice sheets. High spots dominating their surroundings were of prime interest. In E and SE Lithuania, the Baltic Upland hills marking the eastern margin of the last Weichselian glacier hosted numerous fortified settlements from the end of 2nd millennium BC to the Medieval Ages (Narkunai, Velikuskes etc). In W Lithuania, plateau-like hills of the insular Samogitian Upland had been repeatedly fortified from the beginning of 1st millennium AD to the 14th century (Satrija, Medvegalis etc). Chains of hill forts and castle mounds feature the slopes of glaciofluvial valleys of Nemunas, Neris and other rivers where the slopes were dissected by affluent rivulets and ravines and transformed into isolated, well protected hills (Kernave, Punia, Veliuona etc). Peninsulas and headlands formed by the erosion of fluvial and lacustrine deposits were used in the lowlands, e.g. in central and N Lithuania (Paberze, Mezotne etc). How much the landscape was modified for defense purposes? Long-term erosion and overgrowing vegetation damaged the former fortified sites, however some remains and the archeological excavations allowed their reconstruction. The fortified Bronze Age settlements

  19. 76 FR 68625 - Establishment of the Fort Monroe National Monument

    Science.gov (United States)

    2011-11-07

    ... period of slavery in the colonies and, later, this Nation. Two hundred and forty-two years later, Fort... 1863. Thus, Old Point Comfort marks both the beginning and end of slavery in our Nation. The Fort... North Beach area lies the only undeveloped shoreline remaining on Old Point Comfort, providing modern...

  20. Microgrid Enabled Distributed Energy Solutions (MEDES) Fort Bliss Military Reservation

    Science.gov (United States)

    2014-02-01

    FINAL REPORT Microgrid Enabled Distributed Energy Solutions (MEDES) Fort Bliss Military Reservation ESTCP Project EW-201140 FEBRUARY...TITLE AND SUBTITLE Microgrid Enabled Distributed Energy Solutions (MEDES) 5a. CONTRACT NUMBER W912HQ-11-C-0082 Fort Bliss, Texas...Lockheed Martin’s Intelligent Microgrid Solution can provide more energy security while also lowering electric utility costs and greenhouse gas emissions

  1. Fort Valley's early scientists: A legacy of distinction

    Science.gov (United States)

    Andrew J. Sanchez Meador; Susan D. Olberding

    2008-01-01

    When the Riordan brothers of Flagstaff, Arizona, asked Gifford Pinchot to determine why there was a deficit in ponderosa pine seedlings, neither party understood the historical significance of what they were setting in motion for the field of forest research. The direct result of that professional favor was the establishment of the Fort Valley Experiment Station (Fort...

  2. Fort Valley's early scientists: A legacy of distinction (P-53)

    Science.gov (United States)

    Andrew J. Sanchez Meador; Susan D. Olberding

    2008-01-01

    When the Riordan brothers of Flagstaff, Arizona asked Gifford Pinchot to determine why there was a deficit in ponderosa pine seedlings, neither party understood the historical significance of what they were setting in motion for the field of forest research. The direct result of that professional favor was the establishment of the Fort Valley Experiment Station (Fort...

  3. St. Lucia.

    Science.gov (United States)

    1987-06-01

    The population of St Lucia was 123,000 in 1986, with an annual growth rate of 2%. The infant mortality rate stands at 22.2/1000 live births, and life expectancy is 70.3 years for males and 74.9 years for females. The literacy rate is 78%. St Lucia's labor force is allocated as follows: agriculture, 36.6%; industry and commerce, 20.1%; and services, 18.1%. The gross national product (GNP) was US$146 million in 1985, with an annual growth rate of 3% and a per capita GNP of $1071. St Lucia is a parliamentary democracy modeled on the British Westminster system. The island is divided into 16 parishes and 1 urban area (the capital, Castries). St Lucia is currently a politically stable country, although the high level of youth unemployment is a cause for concern. Ongoing stability may depend on the government's ability to provide services such as jobs and housing. The economy has evolved from a monocrop sugar plantation type to a diversified economy based on agriculture, industry, and tourism. Agriculture, dominated by the banana industry, is characterized by the participation of a large number of small and medium-sized enterprises. Industry is being encouraged through the provision of incentives such as tax rebates. The government is attempting to maintain a sound investment climate through a tripartite dialogue with the private sector and trade unions. Overall economic policy is predicated on the attraction of sound investments, by both local and foreign entities, to accelerate the rate of economic growth, solve the unemployment problem, and generate a solid balance-of-payments position.

  4. Research reactors in Austria - Present situation

    International Nuclear Information System (INIS)

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

    2005-01-01

    In the past decades Austria operated three research reactors, the 10 MW ASTRA reactor at Seibersdorf, the 250 kW TRIGA reactor at the Atominstitut and the 1 kW Argonaut reactor at the Technical University in Graz. Since the shut down of the ASTRA on July 31th, 1999 and its immediate decommissioning reactor and the shut down of the Argonaut reactor in Graz on August 31st, 2004 only one reactor remains operational for keeping nuclear competence in Austria which is the 250 kW TRIGA Mark II reactor. (author)

  5. Peter St. John | NREL

    Science.gov (United States)

    St. John Photo of Peter St. John Peter St. John Researcher III-Chemical Engineering Peter.StJohn @nrel.gov | 303-384-7969 Orcid ID http://orcid.org/0000-0002-7928-3722 Education Peter St. John received his engineering from the University of California at Santa Barbara in 2015. During his Ph.D., St. John applied

  6. The Fort Smith radioactive belt, Northwest Territories

    International Nuclear Information System (INIS)

    Charbonneau, B.W.

    1980-01-01

    The Fort Smith Belt is an elongate zone, about 200 km x 50 km, extending from the East Arm of Great Slave Lake southerly into northeastern Alberta. The major feature of the belt is that it is one of the most radioactive regions so far recognized in the Canadian Shield. Potassium, uranium, and thorium are all enriched but the greatest increase is in thorium. The dominant rock type underlying the area is a foliated porphyritic granite. This rock contains an average of about 80 ppm thorium (with areas of tens of square kilometres containing up to 200 ppm) and approximately 11 ppm uranium. In places, dark elongate zones rich in biotite, apatite, and opaque minerals within the porphyritic granite may contain an order of magnitude more uranium and thorium than the porphyry. Radioactive minerals within both the porphyry and the dark zones are principally monazite (containing up to 16% ThO 2 ) and isolated grains of uraninite. This foliated porphyritic granite is interpreted as being pre- or syntectonic with respect to the Hudsonian event because its foliation parallels that of the surrounding rocks. There has been subsequent deformation. The second characteristic feature of the Fort Smith Belt is the development of a peripheral zone where eU is enriched relative to eTh correlating mainly with granitoid rocks which surround the thorium-rich area and wherein ratios of eU/eTh exceed 1:2 (compared to the crustal average of 1:4). Uranium may have moved laterally into this marginal area from the thorium-rich porphyry, possibly in a vapour phase. There is a possibility that concentrations of uranium as well as other metals such as Cu, Mo, Zn, Sn, and W could exist in the porphyry and its margin in appropriate chemical and/or structural traps. The radioactive granite rocks of the Fort Smith Belt are adjacent to uranium-thorium occurrences in the nearby Proterozoic Nonacho sediments but whether or not a genetic relationship exists between the two situations is uncertain. (auth)

  7. Rwanda. Dansk støtte er en god investering

    DEFF Research Database (Denmark)

    Larsen, Josefine Kühnel

    2012-01-01

    Danmark fortsætter militær støtte til Rwanda mens andre lande fryser bistand Rwandas regering har siden 2005 bidraget betydeligt til fredsbevarende operationer. Rwanda har gennem sit styrkebidrag til fredsbevarende operationer potentialet til at spille en væsentlig regional rolle pga. landets...... militære styrke og erfaring med forsoning, konfliktløsning og genopbygning efter krigen og folkemordet i 1994. Aktuelt bidrager Rwanda med 4,577 soldater til fredsbevarende missioner i FN-regi, hvilket betyder, at Rwanda er det land i verden, der leverer det sjette største militære styrkebidrag til FN...

  8. Gynecologic Malignancies Post-LeFort Colpocleisis

    Directory of Open Access Journals (Sweden)

    Rayan Elkattah

    2014-01-01

    Full Text Available Introduction. LeFort colpocleisis (LFC is a safe and effective obliterative surgical option for older women with advanced pelvic organ prolapse who no longer desire coital activity. A major disadvantage is the limited ability to evaluate for post-LFC gynecologic malignancies. Methods. We present the first case of endometrioid ovarian cancer diagnosed after LFC and review all reported gynecologic malignancies post-LFC in the English medical literature. Results. This is the second reported ovarian cancer post-LFC and the first of the endometrioid subtype. A total of nine other gynecologic malignancies post-LFC have been reported in the English medical literature. Conclusions. Gynecologic malignancies post-LFC are rare. We propose a simple 3-step strategy in evaluating post-LFC malignancies.

  9. Master environmental plan for Fort Devens, Massachusetts

    Energy Technology Data Exchange (ETDEWEB)

    Biang, C.A.; Peters, R.W.; Pearl, R.H.; Tsai, S.Y. (Argonne National Lab., IL (United States). Energy Systems Div.)

    1991-11-01

    Argonne National Laboratory has prepared a master environmental plan (MEP) for Fort Devens, Massachusetts, for the US Army Toxic and Hazardous Materials Agency. The MEP is an assessment based on environmental laws and regulations of both the federal government and the Commonwealth of Massachusetts. The MEP assess the physical and environmental status of 58 potential hazardous waste sites, including 54 study areas (SAs) that pose a potential for releasing contamination into the environment and 4 areas of concern (AOCs) that are known to have substantial contamination. For each SA or AOC, this MEP describes the known history and environment, identifies additional data needs, and proposes possible response actions. Most recommended response actions consist of environmental sampling and monitoring and other characterization studies. 74 refs., 63 figs., 50 tabs.

  10. Fort Collins Science Center Ecosystem Dynamics Branch

    Science.gov (United States)

    Wilson, Jim; Melcher, C.; Bowen, Z.

    2009-01-01

    Complex natural resource issues require understanding a web of interactions among ecosystem components that are (1) interdisciplinary, encompassing physical, chemical, and biological processes; (2) spatially complex, involving movements of animals, water, and airborne materials across a range of landscapes and jurisdictions; and (3) temporally complex, occurring over days, weeks, or years, sometimes involving response lags to alteration or exhibiting large natural variation. Scientists in the Ecosystem Dynamics Branch of the U.S. Geological Survey, Fort Collins Science Center, investigate a diversity of these complex natural resource questions at the landscape and systems levels. This Fact Sheet describes the work of the Ecosystems Dynamics Branch, which is focused on energy and land use, climate change and long-term integrated assessments, herbivore-ecosystem interactions, fire and post-fire restoration, and environmental flows and river restoration.

  11. Water quality and hydrology in the Fort Belvoir area, Virginia, 1954-55

    Science.gov (United States)

    Durfor, Charles N.

    1961-01-01

    This report summarizes the results of an investigation of water quality and hydrology in the Fort Belvoir, Va., area for the period August 1954 to September 1955. It summarizes and evaluates information about the water resources of this area that are pertinent to the choice of location and operation of an Army nuclear power reactor. The quantity, quality, nature, and use of the local water that might be affected by the location and operation of a reactor in the area were subjects of investigation. Variations in the quality of the water caused by variation in streamflow, tidal effects, and pollution were important facets of the investigation. During extended periods of low streamflow in the Potomac River (usually in the late summer months), salty water moves upstream from Chesapeake Bay and increases the dissolved solids content of the surface waters adjacent to Fort Belvoir. When the streamflow is low the concentration of dissolved solids in the water near the river bottom exceeds that near the surface. The waters in Gunston Cove usually contain more dissolved oxygen than those in the Potomac River. During the summer, the content of dissolved oxygen in the cove waters frequently exceeds 100 percent of saturation. Surface floats that were released on a flood tide in Gunston Cove moved toward the inner portion of the cove in the same direction as the wind and the tide. The maximum average velocity of these floats was 0.65 feet per second. On an ebb tide, many surface floats that were released in Gunston Cove moved toward the inner portion of the cove in the direction of the wind, in opposition to the direction of the tidal movement. Floats released near the mouth of the cove on the same tide, moved with the tide out of the cove through a narrow pass at the end of a submerged sandbar extending from the Fort Belvoir shoreline. The maximum average velocity of the floats in the pass on this ebb tide was 0.85 feet per second. Measurements of subsurface flow direction

  12. Structural remains at the early mediaeval fort at Raibania, Orissa

    Directory of Open Access Journals (Sweden)

    Bratati Sen

    2013-11-01

    Full Text Available The fortifications of mediaeval India occupy an eminent position in the history of military architecture. The present paper deals with the preliminary study of the structural remains at the early mediaeval fort at Raibania in the district of Balasore in Orissa. The fort was built of stone very loosely kept together. The three-walled fortification interspersed by two consecutive moats, a feature evidenced at Raibania, which is unparallel in the history of ancient and mediaeval forts and fortifications in India. Several other structures like the Jay-Chandi Temple Complex, a huge well, numerous tanks and remains of an ancient bridge add to the uniqueness of the Fort in the entire eastern region.

  13. Fortælling og fortolkning i Jyske Bank

    DEFF Research Database (Denmark)

    Albrechtsen, Charlotte

    Afhandlingen præsenterer en undersøgelse af et konkret eksempel på storytelling brugt som strategisk ledelses- og kommunikationsredskab i en organisations interne kommunikation. Eksemplet er fortællingen "Slaget ved Vejle", som stammer fra Jyske Bank og udgør under afhandlingens case. De overordn......Afhandlingen præsenterer en undersøgelse af et konkret eksempel på storytelling brugt som strategisk ledelses- og kommunikationsredskab i en organisations interne kommunikation. Eksemplet er fortællingen "Slaget ved Vejle", som stammer fra Jyske Bank og udgør under afhandlingens case. De......, at medarbejderne forholder sig reflekteret, nuanceret og kritisk til den strategiske fortælling, og at der er stor diversitet i deres oplevelser, fortolkninger og vurderinger af fortællingen. Desuden ser afhandlingen nærmere på hvad begrebet "storytelling" dækker over, og hvordan der hidtil er forsket i...

  14. Strategic Analysis and Plan for Implementing Telemedicine at Fort Greely

    National Research Council Canada - National Science Library

    Bolton, Karl

    2003-01-01

    .... To best accomplish this, a strategic analysis and business case analysis was conducted. Introspective strategic analysis tools revealed an organization that is capable of supporting a telemedicine program at Fort Greely...

  15. RadNet Air Data From Fort Smith, AR

    Science.gov (United States)

    This page presents radiation air monitoring and air filter analysis data for Fort Smith, AR from EPA's RadNet system. RadNet is a nationwide network of monitoring stations that measure radiation in air, drinking water and precipitation.

  16. Wood-Fired Boiler System Evaluation at Fort Stewart, GA

    National Research Council Canada - National Science Library

    Potts, Noel

    2002-01-01

    Part of the plan to modernize the central energy plant (CEP) at Fort Stewart, GA is focused on the installations wood-fired boiler, which provides steam for heating, cooling, and domestic hot water. The U.S...

  17. Le Fort I Maxillary Advancement Using Distraction Osteogenesis

    OpenAIRE

    Combs, Patrick D.; Harshbarger, Raymond J.

    2014-01-01

    Treatment of maxillary hypoplasia has traditionally involved conventional Le Fort I osteotomies and advancement. Advancements of greater than 10 mm risk significant relapse. This risk is greater in the cleft lip and palate population, whose anatomy and soft tissue scarring from prior procedures contributes to instability of conventional maxillary advancement. Le Fort I advancement with distraction osteogenesis has emerged as viable, stable treatment modality correction of severe maxillary hyp...

  18. Structural remains at the early mediaeval fort at Raibania, Orissa

    OpenAIRE

    Sen, Bratati

    2013-01-01

    The fortifications of mediaeval India occupy an eminent position in the history of military architecture. The present paper deals with the preliminary study of the structural remains at the early mediaeval fort at Raibania in the district of Balasore in Orissa. The fort was built of stone very loosely kept together. The three-walled fortification interspersed by two consecutive moats, a feature evidenced at Raibania, w...

  19. Undervisning mellem fortælling og feedback

    DEFF Research Database (Denmark)

    Andersen, Kirsten Margrethe

    2016-01-01

    Feedback gør det muligt for den enkelte at forstå, hvordan jeg kan blive bedre til det, jeg er ved at lære. Fortællinger gør det muligt for den enkelte at udvide horisonten og derved komme til en forståelse af, hvilke mulige perspektiver der er for at forholde sig til den verden, som fortællingen...

  20. BRAND EQUITY OF LAHORE FORT AS A TOURISM DESTINATION BRAND

    OpenAIRE

    KASHIF, MUHAMMAD; SAMSI, SITI ZAKIAH MELATU; SARIFUDDIN, SYAMSULANG

    2015-01-01

    ABSTRACTStudies that measure the brand equity of destination brands by using the Customer-Based Brand Equity (CBBE) model in a developing country context are scarce. The present study investigates the destination brand equity of the Lahore Fort by employing the CBBE model in a developing country context of Pakistan. Following the positivist tradition, we adopted a survey-based approach to collect data from 237 tourists visiting the Lahore Fort. Data were collected through a questionnaire deve...

  1. FOUR YEARS OF OPERATIONS AND RESULTS WITH FORTE

    International Nuclear Information System (INIS)

    D. ROUSSEL-DUPRE; P. KLINGNER; L. CARLSON; ET AL

    2001-01-01

    The FORTE (Fast Onboard Recording of Transient Events) satellite was launched on 29 August 1997 and has been in continuous operation since that time. FORTE was placed in a nearly circular, 825-km-altitude, 70 degrees inclination orbit by a Pegasus rocket funded by Air Force Space Test Program. The Department of Energy funded the FORTE satellite, which was designed and built at Los Alamos. FORTE's successful launch and engineered robustness were a result of several years of dedicated work by the joint Los Alamos National Laboratory/Sandia National Laboratory project team, led through mission definition, payload and satellite development, and launch by Dr. Stephen Knox. The project is now led by Dr. Abram Jacobson. FORTE carries a suite of instruments, an optical system and a rf system, for the study of lightning and anthropogenic signals. As a result of this effort, new understandings of lightning events have emerged as well as a more complete understanding of the relationship between optical and rf lightning events. This paper will provide an overview of the FORTE satellite and will discuss the on orbit performance of the subsystems

  2. Cancer incidence in Fort Chipewyan, Alberta : 1995-2006

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Y. [Alberta Cancer Board, Edmonton, AB (Canada). Division of Population Health and Information Surveillance

    2009-02-15

    A high number of cases of cholangiocarcinoma, a rare form of bile duct cancer, as well as high rates of other cancers were reported by a physician working in Fort Chipewyan, Alberta in 2006. Concerns were raised by local residents, attributing cancers in their community to environmental contamination from a range of industrial development including the oil sands development, uranium mining and pulp mills. However, an initial review of the Alberta Cancer Registry did not confirm an increased incidence of cancer in Fort Chipewyan. In the summer/fall of 2007, a working group was formed to support the Alberta Cancer Board in doing a cluster investigation based on the guidelines of the United States Centre for Disease Control and Prevention. This report presented an investigation to determine if there was an elevated rate of cholangiocarcinoma in Fort Chipewyan and whether there was an elevated rate of cancers overall in Fort Chipewyan. The report provided background information on the Athabasca oil sands, uranium mining, and Fort Chipewyan as well as previous investigations of cancer incidence in Fort Chipewyan. Study methods were also presented with particular reference to study and comparison populations; cancer classification and inclusion criteria; active case ascertainment and verification; methods of analysis; and ethical approval. Results were also presented. The specific cancers that were discussed were cholangiocarcinoma, leukemia, colon cancer, and cancer in First Nations in Alberta. It was concluded that the observed number of cases of cholangiocarcinoma was within the expected range. 121 refs., 12 tabs., 3 figs., 5 appendices.

  3. Fort Collins Science Center fiscal year 2010 science accomplishments

    Science.gov (United States)

    Wilson, Juliette T.

    2011-01-01

    The scientists and technical professionals at the U.S. Geological Survey (USGS), Fort Collins Science Center (FORT), apply their diverse ecological, socioeconomic, and technological expertise to investigate complicated ecological problems confronting managers of the Nation's biological resources. FORT works closely with U.S. Department of the Interior (DOI) agency scientists, the academic community, other USGS science centers, and many other partners to provide critical information needed to help answer complex natural-resource management questions. In Fiscal Year 2010 (FY10), FORT's scientific and technical professionals conducted ongoing, expanded, and new research vital to the science needs and management goals of DOI, other Federal and State agencies, and nongovernmental organizations in the areas of aquatic systems and fisheries, climate change, data and information integration and management, invasive species, science support, security and technology, status and trends of biological resources (including the socioeconomic aspects), terrestrial and freshwater ecosystems, and wildlife resources, including threatened and endangered species. This report presents selected FORT science accomplishments for FY10 by the specific USGS mission area or science program with which each task is most closely associated, though there is considerable overlap. The report also includes all FORT publications and other products published in FY10, as well as staff accomplishments, appointments, committee assignments, and invited presentations.

  4. Miniplate fixation of Le Fort I osteotomies.

    Science.gov (United States)

    Rosen, H M

    1986-12-01

    The use of rigid, internal, three-dimensional fixation using vitallium bone plates in 28 consecutive Le Fort I osteotomies is presented. A minimum follow-up period of 6 months was required for inclusion in this patient group. Maxillary movements included advancements (17), intrusions (9), lengthenings (5), and retrusions (2). The majority of maxillae were moved in more than one plane of space. Technical details, complications, and relapse potential are discussed. Advantages of rigid plate fixation include marked reductions in the length of intermaxillary fixation with light training elastics only. Immediate postoperative airway problems are thereby eliminated. Six months of follow-up would appear to indicate a low potential for osseous relapse when compared to wire osteosynthesis, regardless of the direction of maxillary movement. The major disadvantage is the decreased ability of postoperative orthodontics to move dento-osseous segments if skeletal occlusal disharmony persists postoperatively. For this reason, close attention to preoperative planning and operative technique is critical for the success of this fixation method.

  5. Reactor Physics Behind the Chernobyl Accident

    International Nuclear Information System (INIS)

    Reisch, F.

    1999-01-01

    There are some fourteen Chernobyl type of power reactors (1000 MWe) in operation at five different sites in Eastern Europe. In Russia; in St. Petersburg (4). in Smolensk (3). and in Kursk (4) in the Ukraine in Chernobyl (l) and in Lithuania in Ignalina (2). The oldest one is west of St. Petersburg and the most powerful one is in Ignalina. The reactors at St. Petersburg and in Lithuania are near to the Baltic sea. An intricate reactor construction was the most important cause of the accident. There were other reasons too: human error. politics and economics

  6. Possible Future SOFC - ST Based Power Plants

    DEFF Research Database (Denmark)

    Rokni, Masoud; Scappin, Fabio

    2009-01-01

    Hybrid systems consisting Solid Oxide Fuel Cell (SOFC) on the top of a Steam Turbine (ST) are investigated. The plants are fired by natural gas. A desulfurization reactor removes the sulfur content in the NG while a pre-reformer break down the heavier hydrocarbons. The pre-treated fuel enters...

  7. 75 FR 41922 - Notice of Intent To Rule on Request To Release Airport Property at Fort Smith Regional Airport...

    Science.gov (United States)

    2010-07-19

    ... To Release Airport Property at Fort Smith Regional Airport, Fort Smith, AR AGENCY: Federal Aviation... rule and invites public comment on the release of land at Fort Smith Regional Airport under the.... John Parker, Airport Director, Fort Smith Regional Airport, at the following address: Fort Smith...

  8. Prismatic Core Coupled Transient Benchmark

    International Nuclear Information System (INIS)

    Ortensi, J.; Pope, M.A.; Strydom, G.; Sen, R.S.; DeHart, M.D.; Gougar, H.D.; Ellis, C.; Baxter, A.; Seker, V.; Downar, T.J.; Vierow, K.; Ivanov, K.

    2011-01-01

    The Prismatic Modular Reactor (PMR) is one of the High Temperature Reactor (HTR) design concepts that have existed for some time. Several prismatic units have operated in the world (DRAGON, Fort St. Vrain, Peach Bottom) and one unit is still in operation (HTTR). The deterministic neutronics and thermal-fluids transient analysis tools and methods currently available for the design and analysis of PMRs have lagged behind the state of the art compared to LWR reactor technologies. This has motivated the development of more accurate and efficient tools for the design and safety evaluations of the PMR. In addition to the work invested in new methods, it is essential to develop appropriate benchmarks to verify and validate the new methods in computer codes. The purpose of this benchmark is to establish a well-defined problem, based on a common given set of data, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events. The benchmark-working group is currently seeking OECD/NEA sponsorship. This benchmark is being pursued and is heavily based on the success of the PBMR-400 exercise.

  9. Fort Collins Science Center: Invasive Species Science

    Science.gov (United States)

    Stohlgren, Tom

    2004-01-01

    Invasive, non-native species of plants, animals, and disease organisms adversely affect the ecosystems they enter. Like "biological wildfires," they can quickly spread, and they affect nearly all terrestrial and aquatic ecosystems. Invasive species have become the greatest environmental challenge of the 21st century in terms of economic, environmental, and human health costs, with an estimated impact in the U.S. of over $138 billion per year. Managers of Department of the Interior and other public and private lands and waters rank invasive species as their top resource management problem.

  10. Thunderstorm and Lightning Studies using the FORTE Optical Lightning System (FORTE/OLS)

    Energy Technology Data Exchange (ETDEWEB)

    Argo, P.; Franz, R.; Green, J.; Guillen, J.L.; Jacobson, A.R.; Kirkland, M.; Knox, S.; Spalding, R.; Suszcynsky, D.M.

    1999-02-01

    Preliminary observations of simultaneous RF and optical emissions from lightning as seen by the FORTE spacecraft are presented. RF/optical pairs of waveforms are routinely collected both as individual lightning events and as sequences of events associated with cloud-to-ground (CG) and intra-cloud (IC) flashes. CG pulses can be distinguished from IC pulses based on the properties of the RF and optical waveforms, but mostly based on the associated RF spectrograms. The RF spectrograms are very similar to previous ground-based VHF observations of lightning and show signatures associated with return strokes, stepped and dart leaders, and attachment processes,. RF emissions are observed to precede the arrival of optical emissions at the satellite by a mean value of 280 microseconds. The dual phenomenology nature of these observations are discussed in terms of their ability to contribute to a satellite-based lightning monitoring mission.

  11. Thunderstorm and Lightning Studies using the FORTE Optical Lightning System (FORTE/OLS)

    International Nuclear Information System (INIS)

    Argo, P.; Franz, R.; Green, J.; Guillen, J.L.; Jacobson, A.R.; Kirkland, M.; Knox, S.; Spalding, R.; Suszcynsky, D.M.

    1999-01-01

    Preliminary observations of simultaneous RF and optical emissions from lightning as seen by the FORTE spacecraft are presented. RF/optical pairs of waveforms are routinely collected both as individual lightning events and as sequences of events associated with cloud-to-ground (CG) and intra-cloud (IC) flashes. CG pulses can be distinguished from IC pulses based on the properties of the RF and optical waveforms, but mostly based on the associated RF spectrograms. The RF spectrograms are very similar to previous ground-based VHF observations of lightning and show signatures associated with return strokes, stepped and dart leaders, and attachment processes,. RF emissions are observed to precede the arrival of optical emissions at the satellite by a mean value of 280 microseconds. The dual phenomenology nature of these observations are discussed in terms of their ability to contribute to a satellite-based lightning monitoring mission

  12. Le Fort I Maxillary Advancement Using Distraction Osteogenesis

    Science.gov (United States)

    Combs, Patrick D.; Harshbarger, Raymond J.

    2014-01-01

    Treatment of maxillary hypoplasia has traditionally involved conventional Le Fort I osteotomies and advancement. Advancements of greater than 10 mm risk significant relapse. This risk is greater in the cleft lip and palate population, whose anatomy and soft tissue scarring from prior procedures contributes to instability of conventional maxillary advancement. Le Fort I advancement with distraction osteogenesis has emerged as viable, stable treatment modality correction of severe maxillary hypoplasia in cleft, syndromic, and noncleft patients. In this article, the authors provide a review of current data and recommendations concerning Le Fort I advancement with distraction osteogenesis. In addition, they outline their technique for treating severe maxillary hypoplasia with distraction osteogenesis using internal devices. PMID:25383054

  13. Targeting Net Zero Energy at Fort Carson: Assessment and Recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, K.; Markel, T.; Simpson, M.; Leahey, J.; Rockenbaugh, C.; Lisell, L.; Burman, K.; Singer, M.

    2011-10-01

    The U.S. Army's Fort Carson installation was selected to serve as a prototype for net zero energy assessment and planning. NREL performed the comprehensive assessment to appraise the potential of Fort Carson to achieve net zero energy status through energy efficiency, renewable energy, and electric vehicle integration. This report summarizes the results of the assessment and provides energy recommendations. This study is part of a larger cross-laboratory effort that also includes an assessment of renewable opportunities at seven other DoD Front Range installations, a microgrid design for Fort Carson critical loads and an assessment of regulatory and market-based barriers to a regional secure smart grid.

  14. Cranial nerve injury after Le Fort I osteotomy.

    Science.gov (United States)

    Kim, J-W; Chin, B-R; Park, H-S; Lee, S-H; Kwon, T-G

    2011-03-01

    A Le Fort I osteotomy is widely used to correct dentofacial deformity because it is a safe and reliable surgical method. Although rare, various complications have been reported in relation to pterygomaxillary separation. Cranial nerve damage is one of the serious complications that can occur after Le Fort I osteotomy. In this report, a 19-year-old man with unilateral cleft lip and palate underwent surgery to correct maxillary hypoplasia, asymmetry and mandibular prognathism. After the Le Fort I maxillary osteotomy, the patient showed multiple cranial nerve damage; an impairment of outward movement of the eye (abducens nerve), decreased vision (optic nerve), and paraesthesia of the frontal and upper cheek area (ophthalmic and maxillary nerve). The damage to the cranial nerve was related to an unexpected sphenoid bone fracture and subsequent trauma in the cavernous sinus during the pterygomaxillary osteotomy. Copyright © 2010 International Association of Oral and Maxillofacial Surgeons. Published by Elsevier Ltd. All rights reserved.

  15. Fort Collins Science Center-Fiscal year 2009 science accomplishments

    Science.gov (United States)

    Wilson, Juliette T.

    2010-01-01

    Public land and natural resource managers in the United States are confronted with increasingly complex decisions that have important ramifications for both ecological and human systems. The scientists and technical professionals at the U.S. Geological Survey Fort Collins Science Center?many of whom are at the forefront of their fields?possess a unique blend of ecological, socioeconomic, and technological expertise. Because of this diverse talent, Fort Collins Science Center staff are able to apply a systems approach to investigating complicated ecological problems in a way that helps answer critical management questions. In addition, the Fort Collins Science Center has a long record of working closely with the academic community through cooperative agreements and other collaborations. The Fort Collins Science Center is deeply engaged with other U.S. Geological Survey science centers and partners throughout the Department of the Interior. As a regular practice, we incorporate the expertise of these partners in providing a full complement of ?the right people? to effectively tackle the multifaceted research problems of today's resource-management world. In Fiscal Year 2009, the Fort Collins Science Center's scientific and technical professionals continued research vital to Department of the Interior's science and management needs. Fort Collins Science Center work also supported the science needs of other Federal and State agencies as well as non-government organizations. Specifically, Fort Collins Science Center research and technical assistance focused on client and partner needs and goals in the areas of biological information management and delivery, enterprise information, fisheries and aquatic systems, invasive species, status and trends of biological resources (including human dimensions), terrestrial ecosystems, and wildlife resources. In the process, Fort Collins Science Center science addressed natural-science information needs identified in the U

  16. Black Swan Event Assessment for Fort Leonard Wood, Missouri

    Science.gov (United States)

    2016-03-01

    ER D C/ CE RL S R- 16 -1 Net Zero Planning for Fort Leonard Wood Black Swan Event Assessment for Fort Leonard Wood, Missouri Co ns...search for other technical reports published by ERDC, visit the ERDC online library at http://acwc.sdp.sirsi.net/client/default. Net Zero Planning for...1.8 degrees Celsius knots 0.5144444 meters per second miles (US statute) 1,609.347 meters miles per hour 0.44704 meters per second ERDC/CERL SR

  17. Sundhedsprofessionelles begejstring for fortællinger fra levet erfaring

    DEFF Research Database (Denmark)

    Liveng, Anne; Larsen, Christine; Lange, Mads

    2018-01-01

    I 2013 etablerede psykiatrien i Region Hovedstaden, Danmark, et undervisningsprogram om recovery for sundhedsprofessionelle. Evalueringer af programmet viste et udtalt engagement i fortællingen fra underviseren med levet erfaring. Artiklen diskuterer hvordan dette kan forstås. Evalueringsmaterialet...... analyseres ud fra et læringsteoretisk perspektiv og fokuserer på: 1) Betydningen af fortællingens emotionelle indhold, 2) Rolle-bytningen mellem personen med levet erfaring og sundhedsprofessionelle, og 3) Workshoppene som et læringsrum, der aktiverer refleksioner over strukturer og organisering af...

  18. Renewable Energy Opportunities at Fort Campbell, Tennessee/Kentucky

    Energy Technology Data Exchange (ETDEWEB)

    Hand, James R.; Horner, Jacob A.; Kora, Angela R.; Orrell, Alice C.; Russo, Bryan J.; Weimar, Mark R.; Nesse, Ronald J.

    2011-03-31

    This document provides an overview of renewable resource potential at Fort Campbell, based primarily upon analysis of secondary data sources supplemented with limited on-site evaluations. This effort focuses on grid-connected generation of electricity from renewable energy sources and also on ground source heat pumps for heating and cooling buildings. The effort was funded by the U.S. Army Installation Management Command (IMCOM) as follow-on to the 2005 Department of Defense (DoD) Renewables Assessment. The site visit to Fort Campbell took place on June 10, 2010.

  19. Renewable Energy Opportunities at Fort Drum, New York

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Scott A.; Orrell, Alice C.; Solana, Amy E.; Williamson, Jennifer L.; Hand, James R.; Russo, Bryan J.; Weimar, Mark R.; Rowley, Steven; Nesse, Ronald J.

    2010-10-20

    This document provides an overview of renewable resource potential at Fort Drum, based primarily upon analysis of secondary data sources supplemented with limited on-site evaluations. This effort focuses on grid-connected generation of electricity from renewable energy sources and also on ground source heat pumps for heating and cooling buildings. The effort was funded by the U.S. Army Installation Management Command (IMCOM) as follow-on to the 2005 Department of Defense (DoD) Renewables Assessment. The site visit to Fort Drum took place on May 4 and 5, 2010.

  20. En fascinerende fortælling om det 20. århundredes musik

    DEFF Research Database (Denmark)

    Bonde, Lars Ole

    2011-01-01

    Anmeldelse af Karl Aage Rasmussen: Musik i det tyvende århundrede: En fortælling. Gyldendal 2011.......Anmeldelse af Karl Aage Rasmussen: Musik i det tyvende århundrede: En fortælling. Gyldendal 2011....

  1. 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

  2. Fort Independence: An Eighteenth-Century Frontier Homesite and Militia Post in South Carolina.

    Science.gov (United States)

    1982-12-01

    included in this instance as a condiment , but it could also indicate that the Fort Independence garrison was familiar with the strategy employed by the Fort...archeological investigation of Fort Charlotte, McCormick County, South Carolina. Notebook, Institute of Archeology and Anthropology, University of South

  3. 40 CFR 81.63 - Metropolitan Fort Smith Interstate Air Quality Control Region.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 17 2010-07-01 2010-07-01 false Metropolitan Fort Smith Interstate Air... Air Quality Control Regions § 81.63 Metropolitan Fort Smith Interstate Air Quality Control Region. The Metropolitan Fort Smith Interstate Air Quality Control Region (Arkansas-Oklahoma) has been revised to consist...

  4. Case Study: Fort Mill High School--A Culture of Continuous Improvement

    Science.gov (United States)

    Southern Regional Education Board (SREB), 2014

    2014-01-01

    This is the latest in a series of case studies highlighting best practices High Schools That Work (HSTW) network schools and districts are implementing to prepare students better for further studies and careers. Fort Mill High School is in Fort Mill, South Carolina, an outlying suburb of Charlotte, North Carolina. Fort Mill links high quality…

  5. Fort Peck-Wolf Point transmission line project, Montana

    International Nuclear Information System (INIS)

    1992-01-01

    The primary objective of the project is to replace the existing 36-mile Fort Peck-Wolf Point transmission line which has reached the end of its useful service life. Presently, the overall condition of this existing section of the 47-year-old line is poor. Frequent repairs have been required because of the absence of overhead ground wires. The continued maintenance of the line will become more expensive and customer interruptions will persist because of the damage due to lightning. The expense of replacing shell rotted poles, and the concern for the safety of the maintenance personnel because of hazards caused by severe shell rot are also of primary importance. The operational and maintenance problems coupled with power system simulation studies, demonstrate the need for improvements to the Wolf Point area to serve area loads. Western's Wolf Point Substation is an important point of interconnection for the power output from the Fort Peck Dam to area loads as far away as Williston, North Dakota. The proposed transmission line replacement would assure that there will continue to be reliable transmission capacity available to serve area electrical loads, as well as provide a reliable second high-voltage transmission path from the Fort Peck generation to back-up a loss of the Fort Peck-Wolf Point 115-kV Line No. 1

  6. Site-Based Budgeting in Fort Worth, Texas.

    Science.gov (United States)

    Peternick, Lauri; Sherman, Joel

    1998-01-01

    Examines the Fort Worth Independent School District's decentralized decision-making system through three lenses: a review of site-based decision-making procedures at several schools; an examination of who participates; and stakeholders' perceptions. Some schools operated democratically, significantly including teachers, parents, and community…

  7. 22nd Spring Research Festival Showcases Fort Detrick Science | Poster

    Science.gov (United States)

    Rainy weather couldn’t dampen the spirits of visitors to the 2018 Spring Research Festival, which brought together scientists from the Frederick National Laboratory (FNL), NCI at Frederick, and the U.S. Army Medical Research and Materiel Command (USAMRMC) and showcased the important research that takes place every day at Fort Detrick.

  8. The health of loblolly pine stands at Fort Benning, GA

    Science.gov (United States)

    Soung-Ryoul Ryu; G. Geoff Wang; Joan L. Walker

    2013-01-01

    Approximately two-thirds of the red-cockaded woodpecker (Picoides borealis) (RCW) groups at Fort Benning, GA, depend on loblolly pine (Pinus taeda) stands for nesting or foraging. However, loblolly pine stands are suspected to decline. Forest managers want to replace loblolly pine with longleaf pine (P. palustris...

  9. Forest pathology and entomology at Fort Valley Experimental Forest

    Science.gov (United States)

    Brian W. Geils

    2008-01-01

    Forest pathology and entomology have been researched at Fort Valley Experimental Forest throughout its history. The pathogens and insects of particular interest are mistletoes, decay and canker fungi, rusts, bark beetles, and various defoliators. Studies on life history, biotic interactions, impacts, and control have been published and incorporated into silvicultural...

  10. Tile forts of the Liesbeeck Frontier | Sleigh | Scientia Militaria: South ...

    African Journals Online (AJOL)

    Scientia Militaria: South African Journal of Military Studies. Journal Home · ABOUT THIS JOURNAL · Advanced Search · Current Issue · Archives · Journal Home > Vol 27 (1997) >. Log in or Register to get access to full text downloads. Username, Password, Remember me, or Register. Tile forts of the Liesbeeck Frontier.

  11. Opening remarks for the Fort Valley Centennial Celebration

    Science.gov (United States)

    G. Sam Foster

    2008-01-01

    The Rocky Mountain Research Station recognizes and values the contributions of our scientists and collaborators for their work over the past century at Fort Valley Experimental Forest. With the help of our partners and collaborators, Rocky Mountain Research Station is working to improve coordination across its research Program Areas and Experimental Forests and Ranges...

  12. Opening remarks for the Fort Valley Centennial Celebration (P-53)

    Science.gov (United States)

    G. Sam Foster

    2008-01-01

    The Rocky Mountain Research Station recognizes and values the contributions of our scientists and collaborators for their work over the past century at Fort Valley Experimental Forest. With the help of our partners and collaborators, Rocky Mountain Research Station is working to improve coordination across its research Program Areas and Experimental Forests and Ranges...

  13. Informal report on measurements of slant TEC by FORTE

    International Nuclear Information System (INIS)

    Massey, R.S.

    1997-01-01

    Los Alamos National Laboratory's Space and Atmospheric Sciences group is now operating the FORTE satellite, which has two sets of instruments: optical detectors and radio detectors. In this report the author describes work with one set of radio detectors that allow measurements of the total electron content (TEC) traversed by VHF radiation originating at an electromagnetic pulse (EMP) generator located at Los Alamos

  14. Assessment of DoD Wounded Warrior Matters -- Fort Riley

    Science.gov (United States)

    2013-08-06

    acceptable excuses included At Remote Care, Regular Leave, Maternity and Paternity Leave, Terminal Leave, Permanent Change of Station, and Transferred to...risk of negative medication interactions and reactions for Soldiers assigned to the Fort Riley WTB. B.2. Background The Joint Commission, an...reconciliation is to minimize medication errors such as omissions, duplications, dosing errors, and drug interactions . Medical reconciliation should

  15. The Forte Kreis : an Attempt to Spiritual Leadership over Europe

    NARCIS (Netherlands)

    Poorthuis, Marcel

    2017-01-01

    Just before the outbreak of World War 1, a group of writers, artists and philosophers decided to establish a spiritual rule over Europe, the Forte Kreis. The group aimed at a reconciliation in Europe, by establishing pacifism, but also between East and West by creating a new language. Their thoughts

  16. Early thinning experiments established by the Fort Valley Experimental Forest

    Science.gov (United States)

    Benjamin P. De Blois; Alex. J. Finkral; Andrew J. Sanchez Meador; Margaret M. Moore

    2008-01-01

    Between 1925 and 1936, the Fort Valley Experimental Forest (FVEF) scientists initiated a study to examine a series of forest thinning experiments in second growth ponderosa pine stands in Arizona and New Mexico. These early thinning plots furnished much of the early background for the development of methods used in forest management in the Southwest. The plots ranged...

  17. 77 FR 24579 - Establishment of the Fort Ord National Monument

    Science.gov (United States)

    2012-04-25

    ... 1775-1776, Anza established the first overland route from ``New Spain,'' as Mexico was then known, to..., approximately 6 miles of which pass through the Fort Ord area. Although much of the historic route currently... tourists and recreationalists from near and far, and enhance its unique natural resources, for the...

  18. Fort Valley studies: A natural laboratory for research and education

    Science.gov (United States)

    Brian W. Geils

    2008-01-01

    Drought, wildfire, extinction, and invasive species are considered serious threats to the health of our forests. Although these issues have global connections, we most readily see their consequences locally and attempt to respond with management based on science. For 100 years, the Fort Valley Experimental Forest (FVEF) has provided educational and experimental support...

  19. Characteristics of potential repository wastes

    International Nuclear Information System (INIS)

    1992-07-01

    The LWR spent fuels discussed in Volume 1 of this report comprise about 99% of all domestic non-reprocessed spent fuel. In this report we discuss other types of spent fuels which, although small in relative quantity, consist of a number of diverse types, sizes, and compositions. Many of these fuels are candidates for repository disposal. Some non-LWR spent fuels are currently reprocessed or are scheduled for reprocessing in DOE facilities at the Savannah River Site, Hanford Site, and the Idaho National Engineering Laboratory. It appears likely that the reprocessing of fuels that have been reprocessed in the past will continue and that the resulting high-level wastes will become part of defense HLW. However, it is not entirely clear in some cases whether a given fuel will be reprocessed, especially in cases where pretreatment may be needed before reprocessing, or where the enrichment is not high enough to make reprocessing attractive. Some fuels may be canistered, while others may require special means of disposal. The major categories covered in this chapter include HTGR spent fuel from the Fort St. Vrain and Peach Bottom-1 reactors, research and test reactor fuels, and miscellaneous fuels, and wastes generated from the decommissioning of facilities

  20. Characteristics of potential repository wastes. Volume 2

    Energy Technology Data Exchange (ETDEWEB)

    1992-07-01

    The LWR spent fuels discussed in Volume 1 of this report comprise about 99% of all domestic non-reprocessed spent fuel. In this report we discuss other types of spent fuels which, although small in relative quantity, consist of a number of diverse types, sizes, and compositions. Many of these fuels are candidates for repository disposal. Some non-LWR spent fuels are currently reprocessed or are scheduled for reprocessing in DOE facilities at the Savannah River Site, Hanford Site, and the Idaho National Engineering Laboratory. It appears likely that the reprocessing of fuels that have been reprocessed in the past will continue and that the resulting high-level wastes will become part of defense HLW. However, it is not entirely clear in some cases whether a given fuel will be reprocessed, especially in cases where pretreatment may be needed before reprocessing, or where the enrichment is not high enough to make reprocessing attractive. Some fuels may be canistered, while others may require special means of disposal. The major categories covered in this chapter include HTGR spent fuel from the Fort St. Vrain and Peach Bottom-1 reactors, research and test reactor fuels, and miscellaneous fuels, and wastes generated from the decommissioning of facilities.

  1. St. John's Wort (image)

    Science.gov (United States)

    The herb St. John's Wort is believed to be helpful in relieving mild to moderate depression, but should only be taken under a physician's supervision. St. John's Wort may clash with other medications or ...

  2. Morphological anomalies in two Lutzomyia (Psathyromyia) shannoni (Diptera: Psychodidae: Phlebotominae) specimens collected from Fort Rucker, Alabama, and Fort Campbell, Kentucky.

    Science.gov (United States)

    Florin, David A; Lawyer, Phillip; Rowton, Edgar; Schultz, George; Wilkerson, Richard; Davies, Stephen J; Lipnick, Robert; Keep, Lisa

    2010-09-01

    This report describes two male specimens of the sand fly species Lutzomyia shannoni (Dyar) (Diptera: Psychodidae: Phlebotominae) collected at Fort Rucker, AL, and Fort Campbell, KY, in dry ice-baited light traps during September 2005. The specimens were observed to have anomalies to the number of spines on the gonostyli. The taxonomic keys of Young and Perkins (Mosq. News 44: 263-285; 1984) use the number of spines on the gonostylus in the first couplet to differentiate two major groupings of North American sand flies. The two anomalous specimens were identified as L. shannoni based on the following criteria: (1) both specimens possess antennal ascoids with long, distinct proximal spurs (a near diagnostic character of L. shannoni in North America), (2) the sequences of the partial cytochrome c oxidase subunit 1 gene from both specimens indicated L. shannoni, and (3) the sequences of the internal transcribed spacer 2 molecular marker from both specimens indicated L. shannoni. The anomalous features are fundamentally different from each other as the Fort Rucker specimen possesses a fifth spine (basally located) on just one gonostylus, whereas the Fort Campbell specimen possesses five spines (extra spines subterminally located) on both gonostyli. Because the gonostyli are part of the external male genitalia, anomalies in the number of spines on the gonostyli may have serious biological consequences, such as reduced reproductive success, for the possessors. These anomalies are of taxonomic interest as the specimens could easily have been misidentified using available morphological keys.

  3. Generic magnetic fusion reactor cost assessment

    International Nuclear Information System (INIS)

    Sheffield, J.

    1985-01-01

    The Fusion Energy Division of the Oak Ridge National Laboratory discusses ''generic'' magnetic fusion reactors. The author comments on DT burning magnetic fusion reactor models being possibly operational in the 21st century. Representative parameters from D-T reactor studies are given, as well as a shematic diagram of a generic fusion reactor. Values are given for winding pack current density for existing and future superconducting coils. Topics included are the variation of the cost of electricity (COE), the dependence of the COE on the net electric power of the reactor, and COE formula definitions

  4. 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...

  5. Broad-Application Test Reactor

    International Nuclear Information System (INIS)

    Motloch, C.G.

    1992-05-01

    This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

  6. Master Environmental Plan: Fort Wingate Depot Activity, Gallup, New Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Biang, C.A.; Yuen, C.R.; Biang, R.P.; Antonopoulos, A.A.; Ditmars, J.D.

    1990-12-01

    The master environmental plan is based on an environmental assessment of the areas requiring environmental evaluation (AREEs) at Fort Wingate Depot Activity near Gallup, New Mexico. The Fort Wingate Depot Activity is slated for closure under the Base Closure and Realignment Act, Public Law 100--526. The MEP assesses the current status, describes additional data requirements, recommends actions for the sites, and establishes a priority order for actions. The plan was developed so that actions comply with hazardous waste and water quality regulations of the State of New Mexico and applicable federal regulations. It contains a brief history of the site, relevant geological and hydrological information, and a description of the current status for each AREE along with a discussion of the available site-specific data that pertain to existing or potential contamination and the impact on the environment. 35 refs., 27 figs., 23 tabs.

  7. Bastion on the Border: Fort Bliss, 1854-1943

    Science.gov (United States)

    1993-01-01

    Journals , Record Group (RG) 165, Records of the War Department General and Special Staffs, National Archives, Washington, D.C. "’Richard Estrada, Border...it was the Fort Bliss garrison and the other troops deployed by Steever in the El Paso Patrol District that would have to provide the figurative glue ...little military value. For example, on March 30 three carloads of oats, flour , corn, and hay were dispatched; on April 7 fourteen carloads of hay, gasoline

  8. Basewide Energy Study, Fort Wainwright Alaska: Volume 1-Executive Summary

    Science.gov (United States)

    1982-04-01

    more accurate condensate wiett?Ing. 2.2 ENERGY OSAGE ANALISIS 4 top~down anay2ts was mad" of FY’•0 ener;’ uxage’ t Fort wrtsvrigFnt. The spporiIonments...vice, at each receptacle cluster . It should be thermally sensitlve. rtdtcing through-put from 600 watts at -SOOT to0soer power at 100? outside air

  9. Die opleiding van bedryfsielkundiges aan die universiteit van Fort Hare

    Directory of Open Access Journals (Sweden)

    W. Botha

    1977-11-01

    Full Text Available Die Departement Bedryfsielkunde aan die Universiteit van Fort Hare is 'n relatiewe jong departement en het eers in 1965 tot stand gekom. Voor hierdie datum is Bedryfsielkunde as 'n kort kursus deur die departement van suiwer Sielkunde aangebied en een van die destydse dosente, Dr. W. Backer, het die inisiatief geneem om 'n selfstandige departement van Bedryfsielkunde in die Fakulteit van Ekonomiese Wetenskappe op die been te bring.

  10. Assessment of DOD Wounded Warrior Matters -- Fort Drum

    Science.gov (United States)

    2011-09-30

    their relation to military duties. The six factors that are evaluated are: physical capacity or stamina , upper extremities, lower extremities...Health Net Federal Services contractor. The Fort Drum MEDDAC Referral Management Office created a “Reports Cell ” which was responsible for obtaining...Care Division had created a CLR/Reports Cell group that focused specifically on obtaining CLRs, inputting them into patients’ AHLTA records and

  11. Liquid radwaste processing history at Fort Calhoun Nuclear Station

    International Nuclear Information System (INIS)

    Bilau, A.; Rutar, F.

    1989-01-01

    This report presents a historical perspective of liquid radwaste processing at the Fort Calhoun Unit 1 Nuclear Power Station, located in east central Nebraska. Of particular interest is the textual and graphical comparison of the operational implications of the various waste processing methods employed in the last ten years at the Fort Calhoun Station. Fort Calhoun's waste collection and processing systems are described in detail. These process systems include evaporation and solidification employing an in-plant drum solidification system. This solidification system was later replaced with vendor solidification services which solidified wastes in large liners. Ultimately, the plant converted its processing operation to ion exchange cleanup using ion selective media. The operational and economic impact of each of these process systems is discussed including overall costs, personnel exposure, capital expenditure requirements, burial volumes generated, maintenance and reliability assessments. Operational goals and performance criteria employed in the decision-making process for selection of the optimal technology are discussed, including the impact of various influent and effluent requirements

  12. BRAND EQUITY OF LAHORE FORT AS A TOURISM DESTINATION BRAND

    Directory of Open Access Journals (Sweden)

    Muhammad Kashif

    2015-06-01

    Full Text Available Studies that measure the brand equity of destination brands by using the Customer-Based Brand Equity (CBBE model in a developing country context are scarce. The present study investigates the destination brand equity of the Lahore Fort by employing the CBBE model in a developing country context of Pakistan. Following the positivist tradition, we adopted a survey-based approach to collect data from 237 tourists visiting the Lahore Fort. Data were collected through a questionnaire developed to explain the relationship of brand awareness, brand image, brand association, and brand loyalty with Lahore Fort’s overall brand equity. We used various robust statistical techniques such as correlation, regression and confirmatory factor analysis (using PLS method to reach meaningful conclusions and found that brand image and brand associations positively contribute to brand loyalty. Furthermore, brand loyalty significantly contributes towards overall brand equity. Pragmatically, this study measures the customer based brand equity of the Lahore Fort, a destination brand. The results are useful as they suggest a few strategies that can help policy makers to enhance Lahore Fort’s brand performance.

  13. National Training Center Fort Irwin expansion area aquatic resources survey

    Energy Technology Data Exchange (ETDEWEB)

    Cushing, C.E.; Mueller, R.P.

    1996-02-01

    Biologists from Pacific Northwest National Laboratory (PNNL) were requested by personnel from Fort Irwin to conduct a biological reconnaissance of the Avawatz Mountains northeast of Fort Irwin, an area for proposed expansion of the Fort. Surveys of vegetation, small mammals, birds, reptiles, amphibians, and aquatic resources were conducted during 1995 to characterize the populations and habitats present with emphasis on determining the presence of any species of special concern. This report presents a description of the sites sampled, a list of the organisms found and identified, and a discussion of relative abundance. Taxonomic identifications were done to the lowest level possible commensurate with determining the status of the taxa relative to its possible listing as a threatened, endangered, or candidate species. Consultation with taxonomic experts was undertaken for the Coleoptera ahd Hemiptera. In addition to listing the macroinvertebrates found, the authors also present a discussion related to the possible presence of any threatened or endangered species or species of concern found in Sheep Creek Springs, Tin Cabin Springs, and the Amargosa River.

  14. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    International Nuclear Information System (INIS)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

  15. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  16. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  17. An aerial radiological survey of the Fort Calhoun Nuclear Power Plant and surrounding area, Fort Calhoun, Nebraska

    International Nuclear Information System (INIS)

    1994-05-01

    An aerial radiological survey was conducted over the Fort Calhoun Nuclear Power Plant in Fort Calhoun, Nebraska, during the period June 19 through June 28, 1993. The survey was conducted at an altitude of 150 feet (46 meters) over a 25-square-mile (65-square-kilometer) area centered on the power station. The purpose of the survey was to document the terrestrial gamma radiation environment of the Fort Calhoun Nuclear Power Plant and surrounding area. The results of the aerial survey are reported as inferred gamma radiation exposure rates at 1 meter above ground level in the form of a contour map. Outside the plant boundary, exposure rates were found to vary between 6 and 12 microroentgens per hour and were attributed to naturally-occurring uranium, thorium, and potassium. The aerial data were compared to ground-based benchmark exposure rate measurements and radionuclide assays of soil samples obtained within the survey boundary. The ground-based measurements were found to be in good agreement with those inferred from the aerial measuring system. A previous survey was conducted on August 9 and 10, 1972, before the plant began operation. Exposure rates measured in both surveys were consistent with normal terrestrial background

  18. Reactor Physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, A.

    2002-01-01

    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

  19. 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.

  20. 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.

  1. 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

  2. Properties of unirradiated fuel element graphites H-451 and SO818. [Bulk density, tensile properties, thermal expansion, thermal conductivity

    Energy Technology Data Exchange (ETDEWEB)

    Engle, G.B.; Johnson, W.R.

    1976-10-08

    Nuclear graphites H-451, lot 440 (Great Lakes Carbon Corporation (GLCC)), and SO818 (Airco Speer Division, Air Reduction Corporation (AS)) are described, and physical, mechanical, and chemical property data are presented for the graphites in the unirradiated state. A summary of the mean values of the property data and of data on TS-1240 and H-451, lot 426, is tabulated. A direct comparison of H-451, lot 426, chosen for Fort St. Vrain (FSV) fuel reload production, TS-1240, and SO818 may be made from the table. (auth)

  3. The geology and mechanics of formation of the Fort Rock Dome, Yavapai County, Arizona

    Science.gov (United States)

    Fuis, Gary S.

    1996-01-01

    The Fort Rock Dome, a craterlike structure in northern Arizona, is the erosional product of a circular domal uplift associated with a Precambrian shear zone exposed within the crater and with Tertiary volcanism. A section of Precambrian to Quaternary rocks is described, and two Tertiary units, the Crater Pasture Formation and the Fort Rock Creek Rhyodacite, are named. A mathematical model of the doming process is developed that is consistent with the history of the Fort Rock Dome.

  4. Analysis of Delayed Sea Breeze Onset for Fort Ord Prescribed Burning Operations

    Science.gov (United States)

    2015-12-01

    DELAYED SEA BREEZE ONSET FOR FORT ORD PRESCRIBED BURNING OPERATIONS by Dustin D. Hocking December 2015 Thesis Advisor: Wendell Nuss Second...AND DATES COVERED Master’s thesis 4. TITLE AND SUBTITLE ANALYSIS OF DELAYED SEA BREEZE ONSET FOR FORT ORD PRESCRIBED BURNING OPERATIONS 5...release; distribution is unlimited 12b. DISTRIBUTION CODE 13. ABSTRACT (maximum 200 words) The U.S. Army conducts prescribed burns at Fort Ord

  5. 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

  6. 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

  7. Surveys in 1961 on St. Thomas & St. Croix

    DEFF Research Database (Denmark)

    Dahl, Thorkel; Licht, Kjeld de Fine

    Registration of towns and buildings erected during the Danish reign of the Caribbean Isles of St. Thomas, St. Jan and St. Croix 1671-1917 (now belonging to the USA under the name of Virgin Islands)....

  8. Process development report: 0.20-m primary burner system

    International Nuclear Information System (INIS)

    Rickman, W.S.

    1978-09-01

    HTGR reprocessing consists of crushing the spent fuel elements to a size suitable for burning in a fluidized bed to remove excess graphite, separating the fissile and fertile particles, crushing and burning the SiC-coated fuel particles to remove the remainder of the carbon, dissolution and separation of the particles from insoluble materials, and solvent extraction separation of the dissolved uranium and thorium. Burning the crushed fuel elements is accomplished in a primary burner. This is a batch-continuous, fluidized-bed process utilizing above-bed gravity fines recycle. In gas-solid separation, a combination of a cyclone and porous metal filters is used. This report documents operational tests performed on a 0.20-m primary burner using crushed fuel representative of both Fort St. Vrain and large high-temperature gas-cooled reactor cores. The burner was reconstructed to a gravity fines recycle mode prior to beginning these tests. Results of two separate and successful 48-hour burner runs and several short-term runs have indicated the operability of this concept. Recommendations are made for future work

  9. Process development report: 0.20-m secondary burner system

    International Nuclear Information System (INIS)

    Rickman, W.S.

    1977-09-01

    HTGR fuel reprocessing consists of crushing the spent fuel elements to a size suitable for burning in a fluidized bed to remove excess graphite; separating, crushing, and reburning the fuel particles to remove the remainder of the burnable carbon; dissolution and separation of the particles from insoluble materials; and solvent extraction separation of the dissolved uranium and thorium. Burning the crushed fuel particles is accomplished in a secondary burner. This is a batch fluidized-bed reactor with in-vessel, off-gas filtration. Process heat is provided by an induction heater. This report documents operational tests performed on a commercial size 0.20-m secondary burner using crushed Fort St. Vrain type TRISO fuel particles. Analysis of a parametric study of burner process variables led to recommending lower bed superficial velocity (0.8 m/s), lower ignition temperature (600 0 C), lower fluid bed operating temperature (850 0 C), lower filter blowback frequency (1 cycle/minute), and a lower fluid bed superficial velocity during final bed burnout

  10. Strengths, weaknesses, opportunities and threats for HTR deployment in Europe

    International Nuclear Information System (INIS)

    Bredimas, Alexandre; Kugeler, Kurt; Fütterer, Michael A.

    2014-01-01

    High temperature nuclear reactors are a technology, of which early versions were demonstrated in the 1960s–1980s in Germany (AVR, THTR) and the United States (Peach Bottom, Fort St. Vrain). HTRs were initially designed for high temperature, high efficiency electricity generation but the technology, the market and the targeted applications have evolved since then to address industrial cogeneration and new operational conditions (in particular new safety regulations). This paper intends to analyse the latest status of HTR today, as regards their intrinsic strengths and weaknesses and their external context, whether positive (opportunities) or negative (threats). Different dimensions are covered by the analysis: technology status, results from R and D programmes (especially in Europe), competences and skills, licensing aspects, experience feedback from demonstrator operation (in particular in Germany), economic conditions and other non-technical aspects. Europe has a comprehensive experience in the field of HTR with capabilities in both pebble bed and prismatic designs (R and D, engineering, manufacturing, operation, dismantling, and the full fuel cycle). Europe is also a promising market for HTR as the process heat market is large with good industrial and cogeneration infrastructures. The analysis of the European situation is to a good deal indicative for the global potential of this technology

  11. Feasibility of monitoring the strength of HTGR core support graphite: Part III

    International Nuclear Information System (INIS)

    Morgan, W.C.; Davis, T.J.; Thomas, M.T.

    1983-02-01

    Methods are being developed to monitor, in-situ, the strength changes of graphite core-support components in a High-Temperature Gas-Cooled Reactor (HTGR). The results reported herein pertain to the development of techniques for monitoring the core-support blocks; the PGX graphite used in these studies is the grade used for the core-support blocks of the Fort St. Vrain HTGR, and is coarser-grained than the grades used in our previous investigations. The through-transmission ultrasonic velocity technique, developed for monitoring strength of the core-support posts, is not suitable for use on the core-support blocks. Eddy-current and ultrasonic backscattering techniques have been shown to be capable of measuring the density-depth profile in oxidized PGX and, combined with a correlation of strength versus density, could yield an estimate of the strength-depth profile of in-service HTGR core support blocks. Correlations of strength versus density and other properties, and progress on the development of the eddy-current and ultrasonic backscattering techniques are reported

  12. Analytic solutions to linear, time-dependent fission product deposition models for isothermal laminar, slug, or multiregion flow conditions

    International Nuclear Information System (INIS)

    Durkee, J.W. Jr.

    1983-01-01

    The time-dependent convective-diffusion equation with radioactive decay is solved analytically in axisymmetric cylindrical geometry for laminar and slug velocity profiles under isothermal conditions. Concentration dependent diffusion is neglected. The laminar flow solution is derived using the method of separation of variables and Frobenius' technique for constructing a series expansion about a regular singular point. The slug flow multiregion solution is obtained using the method of separation of variables. The Davidon Variable Metric Minimization algorithm is used to compute the coupling coefficients. These solutions, which describe the transport of fission products in a flowing stream, are then used to determine the concentration of radioactive material deposited on a conduit wall using a standard mass transfer model. Fission product deposition measurements for five diffusion tubes in a Fort St. Vrain High-Temperature Gas-Cooled reactor plateout probe are analyzed. Using single region slug and laminar models, the wall mass transfer coefficients, diffusion coefficients, and inlet concentrations are determined using least squares analysis. The diffusion coefficients and inlet concentrations are consistent between tubes. The derived diffusion coefficients and wall mass transfer coefficients are in relative agreement with known literature values

  13. Helium compressor aerodynamic design considerations for MHTGR circulators

    International Nuclear Information System (INIS)

    McDonald, C.F.

    1988-01-01

    Compressor aerodynamic design considerations for both the main and shutdown cooling circulators in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) plant are addressed in this paper. A major selection topic relates to the impeller type (i.e., axial or radial flow), and the aerothermal studies leading to the selection of optimum parameters are discussed. For the conceptual designs of the main and shutdown cooling circulators, compressor blading geometries were established and helium gas flow paths defined. Both circulators are conservative by industrial standards in terms of aerodynamic and structural loading, and the blade tip speeds are particularly modest. Performance characteristics are presented, and the designs embody margin to ensure that pressure-rise growth potential can be accomodated should the circuit resistance possibly increase as the plant design advances. The axial flow impeller for the main circulator is very similar to the Fort St. Vrain (FSV) helium compressor which performs well. A significant technology base exists for the MHTGR plant circulators, and this is highlighted in the paper. (author). 15 refs, 16 figs, 12 tabs

  14. Automated Environmental Data Collection at Fort Benning, Georgia, from May 1999 to July 2001

    National Research Council Canada - National Science Library

    Hahn, Charles

    2002-01-01

    The Department of Defense, Strategic Environmental Research and Development Program, Ecosystem Management Project, Ecosystem Characterization and Monitoring initiative Program at Fort Benning, Georgia...

  15. Fast breeder reactors

    International Nuclear Information System (INIS)

    Ollier, J.L.

    1987-01-01

    The first industrial-scale fast breeder reactor (FBR) is the Superphenix I at Crays-Melville. It was designed and built by Novatome, a French company, and Ansaldo, an Italian company. The advantages of FBRs are summarized. The status of Superphenix and the testing schedule is given. The stages in its power escalation in 1986 are given. The article is optimistic about the future for FBRs and expects FBRs to take over from PWRs at the beginning of the 21st Century. To achieve economic viability, European financial cooperation for the research and development programme is advocated. (UK)

  16. 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.)

  17. Reactor decommissioning

    International Nuclear Information System (INIS)

    Lawton, H.

    1984-01-01

    A pioneering project on the decommissioning of the Windscale Advanced Gas-cooled Reactor, by the UKAEA, is described. Reactor data; policy; waste management; remote handling equipment; development; and recording and timescales, are all briefly discussed. (U.K.)

  18. RA Reactor

    International Nuclear Information System (INIS)

    1978-02-01

    In addition to basic characteristics of the RA reactor, organizational scheme and financial incentives, this document covers describes the state of the reactor components after 18 years of operation, problems concerned with obtaining the licence for operation with 80% fuel, problems of spent fuel storage in the storage pool of the reactor building and the need for renewal of reactor equipment, first of all instrumentation [sr

  19. Multiregion reactors

    International Nuclear Information System (INIS)

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

    1979-08-01

    The study of reflected reactors can be done employing the multigroup diffusion method. The neutron conservation equations, inside the intervals, can be written by fluxes and group constants. A reflected reactor (one and two groups) for a slab geometry is studied, aplying the continuity of flux and current in the interface. At the end, the appropriated solutions for a infinite cylindrical reactor and for a spherical reactor are presented. (Author) [pt

  20. Proceedings of the INAC 2007 International nuclear atlantic conference. Nuclear energy and energetic challenges for 21st. century. 15. Brazilian national meeting on reactor physics and thermal hydraulics; 8. Brazilian national meeting on nuclear applications

    International Nuclear Information System (INIS)

    2007-01-01

    Following the success of INAC 2002 and 2005, INAC 2007 has been held with the 15th Meeting on Reactor Physics and Thermal Hydraulics (XV ENFIR) and the 8th Meeting on Nuclear Applications (VIII ENAN). A key goal of these joint meetings is to bring together scientists to exchange the latest research and development (R and D) information in nuclear science and technology. In the INAC 2007 technical program, plenary sessions, such as round table discussions and keynote lectures, has held to present to the general public the recent advances of peaceful nuclear energy usage, reducing the global warming. Besides, INAC 2007 has offered a poster technical session on Management Systems for Nuclear Organizations. The XV ENFIR has covered all aspects of interdisciplinary R and D related to nuclear reactors, and the VIII ENAN has offered a forum for discussion on nuclear applications in industry, geology, agriculture, medicine, biology, environmental sciences always highlighting the social, economical and environmental improvements generated by nuclear techniques. Both ENFIR and ENAN have also organized oral and poster technical sessions

  1. Proceedings of the INAC 2007 International nuclear atlantic conference. Nuclear energy and energetic challenges for 21st. century. 15. Brazilian national meeting on reactor physics and thermal hydraulics; 8. Brazilian national meeting on nuclear applications

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-07-01

    Following the success of INAC 2002 and 2005, INAC 2007 has been held with the 15th Meeting on Reactor Physics and Thermal Hydraulics (XV ENFIR) and the 8th Meeting on Nuclear Applications (VIII ENAN). A key goal of these joint meetings is to bring together scientists to exchange the latest research and development (R and D) information in nuclear science and technology. In the INAC 2007 technical program, plenary sessions, such as round table discussions and keynote lectures, has held to present to the general public the recent advances of peaceful nuclear energy usage, reducing the global warming. Besides, INAC 2007 has offered a poster technical session on Management Systems for Nuclear Organizations. The XV ENFIR has covered all aspects of interdisciplinary R and D related to nuclear reactors, and the VIII ENAN has offered a forum for discussion on nuclear applications in industry, geology, agriculture, medicine, biology, environmental sciences always highlighting the social, economical and environmental improvements generated by nuclear techniques. Both ENFIR and ENAN have also organized oral and poster technical sessions.

  2. Study of neutral composition of lower thermosphere at Fort Churchill.

    Science.gov (United States)

    Nier, A. O.; Hickman, D. R.

    1973-01-01

    On Feb. 4 and 6, 1969, and May 11, 1970, Aerobee rockets carrying neutral mass spectrometers were flown at Fort Churchill, Canada during conditions of low geomagnetic activity. As in earlier flights at White Sands, New Mexico, each rocket carried both 'open' and 'closed' ion source instruments. Vertical profiles of N2, O2, O, Ar, and He were measured. Results obtained were essentially the same as those observed at White Sands except that for the winter flights helium appeared to be in diffusive equilibrium.

  3. Archeological Testing Fort Hood: 1994-1995. Volume 2

    Science.gov (United States)

    1996-10-01

    ASSOCI TES, INC. (662-22) Archeological Testing at Fort Hood. 1994-199.5 569 -48-1941.1080-134 1935 -058 Figure 7.17 Selected Perforator Types: Awl and...Department of Anthropology, University of Arkansas. Huskey, V. 1935 An Archeological Survey of the Nueces Canyon of Texas, Bulletin of the Texas... epr 064lL.Tan I lms expected 08-FH1 Yellow 4 expected expedctd cd .9 15.Q W n• I less M 0 ~ *~Tax~on Total Total Inmr 53 nac na Vertebra.es 1. FcAuifnm

  4. Fort Carson Building 1860 Biomass Heating Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    Hunsberger, Randolph [National Renewable Energy Lab. (NREL), Golden, CO (United States); Tomberlin, Gregg [National Renewable Energy Lab. (NREL), Golden, CO (United States); Gaul, Chris [National Renewable Energy Lab. (NREL), Golden, CO (United States)

    2015-09-01

    As part of the Army Net-Zero Energy Installation program, the Fort Carson Army Base requested that NREL evaluate the feasibility of adding a biomass boiler to the district heating system served by Building 1860. We have also developed an Excel-spreadsheet-based decision support tool--specific to the historic loads served by Building 1860--with which users can perform what-if analysis on gas costs, biomass costs, and other parameters. For economic reasons, we do not recommend adding a biomass system at this time.

  5. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sato, Morihiko.

    1994-01-01

    Liquid metals such as liquid metal sodium are filled in a reactor container as primary coolants. A plurality of reactor core containers are disposed in a row in the circumferential direction along with the inner circumferential wall of the reactor container. One or a plurality of intermediate coolers are disposed at the inside of an annular row of the reactor core containers. A reactor core constituted with fuel rods and control rods (module reactor core) is contained at the inside of each of the reactor core containers. Each of the intermediate coolers comprises a cylindrical intermediate cooling vessels. The intermediate cooling vessel comprises an intermediate heat exchanger for heat exchange of primary coolants and secondary coolants and recycling pumps for compulsorily recycling primary coolants at the inside thereof. Since a plurality of reactor core containers are thus assembled, a great reactor power can be attained. Further, the module reactor core contained in one reactor core vessel may be small sized, to facilitate the control for the reactor core operation. (I.N.)

  6. Safety of nuclear power plants in the 21st century

    International Nuclear Information System (INIS)

    Kovacs, Z.; Novakova, H.; Rydzi, S.

    2012-01-01

    Discussing the disaster of March 2011 which had a destroying effect on the Fukushima Dai-ichi nuclear power plant, the article presents an overview of the impacts of the earthquake and tsunami on the nuclear power plants in the region, outlines the defence-in-depth concept, and describes the design of the affected BWR type reactors and the accident event sequences leading to the reactor core damage and radioactivity release into the environment. The proposed measures for enhancing nuclear reactor safety in the 21st century are highlighted. (orig.)

  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. 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)

  9. 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

  10. Termination or Transition: A 21st Century Perspective on the Military’s Role in Conflict Resolution

    Science.gov (United States)

    2009-05-01

    Director, Robert F. Baumann, Ph.D. Graduate Degree Programs iii Abstract TRANSITION OR TERMINATION: A 21 ST CENTURY...1992) and James Raymer , In Search of Lasting Results: Military War Termination Doctrine (Fort Leavenworth, KS: US Army Command and General Staff... Robert E. Baumann, and John T. Fishel, Invasion, Intervention, and “Intervasion”: A Concise History of the US Army in Operation Uphold Democracy

  11. St. Croix trap study

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The data set contains detailed information about the catch from 600 trap stations around St. Croix. Data fields include species caught, size data, trap location...

  12. Strategic Energy Management Plan For Fort Buchanan, Puerto Rico

    Energy Technology Data Exchange (ETDEWEB)

    Parker, Steven A.; Hunt, W. D.

    2001-10-31

    This document reports findings and recommendations as a result of a design assistance project with Fort Buchanan with the goals of developing a Strategic Energy Management Plan for the Site. A strategy has been developed with three major elements in mind: 1) development of a strong foundation from which to build, 2) understanding technologies that are available, and 3) exploring financing options to fund the implementation of improvements. The objective of this report is to outline a strategy that can be used by Fort Buchanan to further establish an effective energy management program. Once a strategy is accepted, the next step is to take action. Some of the strategies defined in this Plan may be implemented directly. Other strategies may require the development of a more sophisticated tactical, or operational, plan to detail a roadmap that will lead to successful realization of the goal. Similarly, some strategies are not single events. Rather, some strategies will require continuous efforts to maintain diligence or to change the culture of the Base occupants and their efforts to conserve energy resources.

  13. Neurosensory changes of palatal mucousa following Le Fort I osteotomy

    Directory of Open Access Journals (Sweden)

    Bijan Movahedian Attar

    2009-09-01

    Full Text Available

    • BACKGROUND: This study evaluated the sensation of palatal ucosa before and after Le Fort I osteotomy and compared it based on whether greater palatine nerve has been dissected or not.
    • METHODS: Sixteen patients were studied within one week before  urgery and then one week, 6 weeks, 3 months and 6 months after surgery. Four tests including sharp-blunt discrimination, cold perception, pin prick sensation and electrical stimulation were performed.
    • RESULTS: Mean values of electrical stimulation were significantly higher 6 months after surgery (p < 0.05, on the other hand mean values of pin-prick sensation were significantly lower (p < 0.05. All patients regardless of the condition of greater palatine nerve were responsive to cold perception and sharp-blunt discrimination 6 months after surgery.
    • CONCLUSIONS: Following Le Fort I osteotomy, palatal  esponsiveness to electrical stimulation decreases and mechanical hyper sensitization occurs. Dissection of greater palatine nerve was shown to have no effect on the results.
    • KEYWORDS: Lefort I Osteotomy, Palatal Mocousa, Nerve Recovery.

  14. Fort Stewart integrated resource assessment. Volume 3: Resource assessment

    Energy Technology Data Exchange (ETDEWEB)

    Sullivan, G.P.; Keller, J.M.; Stucky, D.J.; Wahlstrom, R.R.; Larson, L.L.

    1993-10-01

    The US Army Forces Command (FORSCOM) has tasked the US Department of Energy (DOE) Federal Energy Management Program (FEMP), supported by the Pacific Northwest Laboratory, to identify, evaluate, and assist in acquiring all cost-effective energy projects at Fort Stewart. This is part of a model program that PNL is designing to support energy-use decisions in the federal sector. This report provides the results of the fossil fuel and electric energy resource opportunity (ERO) assessments performed by PNL at the FORSCOM Fort Stewart facility located approximately 25 miles southwest of Savannah, Georgia. It is a companion report to Volume 1, Executive Summary, and Volume 2, Baseline Detail. The results of the analyses of EROs are presented in 11 common energy end-use categories (e.g., boilers and furnaces, service hot water, and building lighting). A narrative description of each ERO is provided, along with a table detailing information on the installed cost, energy and dollar savings; impacts on operations and maintenance (O&M); and, when applicable, a discussion of energy supply and demand, energy security, and environmental issues. A description of the evaluation methodologies and technical and cost assumptions is also provided for each ERO. Summary tables present the cost-effectiveness of energy end-use equipment before and after the implementation of each ERO. The tables also present the results of the life-cycle cost (LCC) analysis indicating the net present value (NPV) and savings to investment ratio (SIR) of each ERO.

  15. Information management needs for Fort Calhoun's design basis reconstitution project

    International Nuclear Information System (INIS)

    Beach, D.R.; Erickson, E.A.; Gambhir, S.K.; Parsons, R.D.

    1989-01-01

    While the need for information management is not new to the nuclear industry or Omaha Public Power District (OPPD), the interrelationship among design information, multiple systems, and design basis issues has necessitated the management of this information in new ways. The project team involved in the reconstitution of the design basis for OPPD's Fort Calhoun nuclear station has experienced the need for the developed effective methods for managing the vast amount of interrelated information associated with this effort. This management of information has been necessary to ensure that design basis documents (DBDs) adequately reflect the interrelated nature of component, system, and plant design; are complete and accurate; and are produced and maintained in a cost-effective manner. Fort Calhoun's aggressive design basis reconstitution project began in early 1987. The present scope of the project includes the production of 52 system and plant level DBDs; currently the project is ∼50% complete with DBDs in various stages of completion, from pilot DBDs through DBDs with approved formats, which have been issued for use. The experience in producing these documents has lead to a growing understanding of the special need for information management in each stage of the project. The development of the information tracking and management processes for the various stages of DBD development has proven to be cost-effective and gives a level of assurance that information has been included in the DBDs consistently and accurately

  16. Von Braun Rocket Team at Fort Bliss, Texas

    Science.gov (United States)

    1940-01-01

    The German Rocket Team, also known as the Von Braun Rocket Team, poses for a group photograph at Fort Bliss, Texas. After World War II ended in 1945, Dr. Wernher von Braun led some 120 of his Peenemuende Colleagues, who developed the V-2 rocket for the German military during the War, to the United Sttes under a contract to the U.S. Army Corps as part of Operation Paperclip. During the following five years the team worked on high altitude firings of the captured V-2 rockets at the White Sands Missile Range in New Mexico, and a guided missile development unit at Fort Bliss, Texas. In April 1950, the group was transferred to the Army Ballistic Missile Agency (ABMA) at Redstone Arsenal in Huntsville, Alabama, and continued to work on the development of the guided missiles for the U.S. Army until transferring to a newly established field center of the National Aeronautic and Space Administration (NASA), George C. Marshall Space Flight Center (MSFC).

  17. 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

  18. Comprehensive Inventory and Determinations of Eligibility for Fort Riley Buildings: 1857-1963

    Science.gov (United States)

    2009-09-01

    become fashionable . Stone residences built at Fort Riley after the 1850s all have rock-faced walls and most have contrasting smooth-faced lintels...507 is significant as a wood-framed Folk Victorian cottage. While Building 507 is one of four Folk Victorian buildings at Fort Riley, it possesses a

  19. A visual progression of the Fort Valley Restoration Project treatments using remotely sensed imagery (P-53)

    Science.gov (United States)

    Joseph E. Crouse; Peter Z. Fule

    2008-01-01

    The landscape surrounding the Fort Valley Experimental Forest in northern Arizona has changed dramatically in the past decade due to the Fort Valley Restoration Project, a collaboration between the Greater Flagstaff Forest Partnership, Coconino National Forest, and Rocky Mountain Research Station. Severe wildfires in 1996 sparked community concern to start restoration...

  20. 75 FR 39051 - Desoto Mills LLC, Fort Payne, AL; Notice of Negative Determination Regarding Application for...

    Science.gov (United States)

    2010-07-07

    ... DEPARTMENT OF LABOR Employment and Training Administration [TA-W-73,416] Desoto Mills LLC, Fort... applicable to workers and former workers at Desoto Mills, LLC, a Subsidiary of Fruit of the Loom, Fort Payne... * * * locations outside the Desoto Mills Plant.'' The petitioner compares the situation at this location with...

  1. 77 FR 71636 - Huntington Foam LLC, Fort Smith, AR; Notice of Revised Determination on Reconsideration

    Science.gov (United States)

    2012-12-03

    ... Smith, AR; Notice of Revised Determination on Reconsideration On August 8, 2012, the Department of Labor... workers and former workers of Huntington Foam LLC, Fort Smith, Arkansas (subject firm). The workers are... reconsideration investigation, I determine that workers of Huntington Foam LLC, Fort Smith, Arkansas, who were...

  2. Recent evolution of HTGR instrumentation in the USA

    International Nuclear Information System (INIS)

    Rodriguez, C.

    1982-06-01

    The reactor instrumentation system for the 2240 MW(t) HTGR includes ex-core neutron detectors for automatic nuclear power control, separate ex-core neutron detectors for automatic protection purposes (reactor trip), reactor core outlet thermocouples that measure the temperature of the primary coolant (helium) as it exits the nuclear core, cold helium thermocouples that measure the temperature of the primary coolant as it enters the core, external pressure differential gages that measure primary coolant flow, in-core fission chambers that are utilized to map neutron flux, and ex-core primary coolant moisture monitors. All of these subsystems, except for the in-core flux mapping units, are also part of the Fort St. Vrain HTGR, which has provided significant experience for the design of the new system. In-core flux mapping is not necessary at FSV for normal operation because its relatively small core is fairly ''visible'' from the location of the ex-core instruments. However, temporary in-core fission couples, microphones, and displacement sensors, as well as sensitive ex-core accelerometers were utilized to identify periodic core block lateral movement and measure neutron flux and primary coolant temperatures. A search for in-core sensors to facilitate mapping neutron flux distributions in the larger core of the 2240 MW(t) HTGR has led to the selection of a high temperature fission chamber, which has been tested up to 1000 deg. C at General Atomic. The chamber shows adequate signal to noise ratio and repeatability. Other reactor instruments planned for the 2240 MW(t) are of the FSV type (i.e. thermocouples) or improved versions of the FSV design (i.e. moisture monitors). New concepts such as acoustic thermometers are also being considered

  3. Advanced reactor development: The LMR integral fast reactor program at Argonne

    International Nuclear Information System (INIS)

    Till, C.E.

    1990-01-01

    Reactor technology for the 21st Century must develop with characteristics that can now be seen to be important for the future, quite different from the things when the fundamental materials and design choices for present reactors were made in the 1950s. Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 3 figs

  4. Possible Future SOFC - ST Based Power Plants

    OpenAIRE

    Rokni, Masoud; Scappin, Fabio

    2009-01-01

    Hybrid systems consisting Solid Oxide Fuel Cell (SOFC) on the top of a Steam Turbine (ST) are investigated. The plants are fired by natural gas. A desulfurization reactor removes the sulfur content in the NG while a pre-reformer break down the heavier hydrocarbons. The pre-treated fuel enters then into the anode side of the SOFC. The gases from the SOFC stacks enter into a burner to burn the rest of the fuel. The off-gases now enter into a heat recovery steam generator to produce steam for a ...

  5. 77 FR 9960 - Final Environmental Impact Statement for Extension of F-Line Streetcar Service to Fort Mason...

    Science.gov (United States)

    2012-02-21

    ... Environmental Impact Statement for Extension of F-Line Streetcar Service to Fort Mason Center, San Francisco, CA... Environmental Impact Statement for the Extension of F-Line Streetcar Service to Fort Mason Center, San Francisco... the extension of the historic streetcar F-line from Fisherman's Wharf to the Fort Mason Center, in San...

  6. 78 FR 28622 - Notice of Approval of Record of Decision for Extending F-Line Streetcar Service to Fort Mason...

    Science.gov (United States)

    2013-05-15

    ...] Notice of Approval of Record of Decision for Extending F-Line Streetcar Service to Fort Mason Center... Environmental Impact Statement (Final EIS) for extending the F-Line historic streetcar service to Fort Mason... turnaround terminus at the Fort Mason Center; and installing appurtenant features such as signals, crossings...

  7. 77 FR 58354 - Bend-Fort Rock Ranger District; Oregon; Withdrawal of Notice for Preparation of an Environmental...

    Science.gov (United States)

    2012-09-20

    ...-Fort Rock Ranger District; Oregon; Withdrawal of Notice for Preparation of an Environmental Impact... Administration, USDOT. ACTION: Notice of withdrawal. SUMMARY: The Bend-Fort Rock Ranger District and FHWA are..., Project Leader, Bend- Fort Rock Ranger District, 63095 Deschutes Market Road, Bend, OR 97701, phone 541...

  8. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  9. 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

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  11. Revisiting "Narrow Bipolar Event" intracloud lightning using the FORTE satellite

    Science.gov (United States)

    Jacobson, A. R.; Light, T. E. L.

    2012-02-01

    The lightning stroke called a "Narrow Bipolar Event", or NBE, is an intracloud discharge responsible for significant charge redistribution. The NBE occurs within 10-20 μs, and some associated process emits irregular bursts of intense radio noise, fading at shorter timescales, sporadically during the charge transfer. In previous reports, the NBE has been inferred to be quite different from other forms of lightning strokes, in two ways: First, the NBE has been inferred to be relatively dark (non-luminous) compared to other lightning strokes. Second, the NBE has been inferred to be isolated within the storm, usually not participating in flashes, but when it is in a flash, the NBE has been inferred to be the flash initiator. These two inferences have sufficiently stark implications for NBE physics that they should be subjected to further independent test, with improved statistics. We attempt such a test with both optical and radio data from the FORTE satellite, and with lightning-stroke data from the Los Alamos Sferic Array. We show rigorously that by the metric of triggering the PDD optical photometer aboard the FORTE satellite, NBE discharges are indeed less luminous than ordinary lightning. Referred to an effective isotropic emitter at the cloud top, NBE light output is inferred to be less than ~3 × 108 W. To address isolation of NBEs, we first expand the pool of geolocated intracloud radio recordings, by borrowing geolocations from either the same flash's or the same storm's other recordings. In this manner we generate a pool of ~2 × 105 unique and independent FORTE intracloud radio recordings, whose slant range from the satellite can be inferred. We then use this slant range to calculate the Effective Radiated Power (ERP) at the radio source, in the passband 26-49 MHz. Stratifying the radio recordings by ERP into eight bins, from a lowest bin (140 kW), we document a trend for the radio recordings to become more isolated in time as the ERP increases. The highest

  12. Nuclear reactor

    International Nuclear Information System (INIS)

    Batheja, P.; Huber, R.; Rau, P.

    1985-01-01

    Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

  13. 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

  14. Determination of one spectral index at the argonaut reactor

    International Nuclear Information System (INIS)

    Klawa, R.

    1973-01-01

    One spectral index at the Argonauta Reactor was determined. The Westcott formalism was employed assuming two components: Maxwellian and 1/E. The values of g(T) and s(T) were obtained from the Westcott definitions by means of the Breit - Wigner formula for the cross section. The r and T were determined for one point at the core of Argonauta Reactor. (author)

  15. Analysis of natural gas supply strategies at Fort Drum

    International Nuclear Information System (INIS)

    Stucky, D.J.; Shankle, S.A.; Anderson, D.M.

    1992-07-01

    This analysis investigates strategies for Fort Drum to acquire a reliable natural gas supply while reducing its gas supply costs. The purpose of this study is to recommend an optimal supply mix based on the life-cycle costs of each strategy analyzed. In particular, this study is intended to provide initial guidance as to whether or not the building and operating of a propane-air mixing station is a feasible alternative to the current gas acquisition strategy. The analysis proceeded by defining the components of supply (gas purchase, gas transport, supplemental fuel supply); identifying alternative options for each supply component; constructing gas supply strategies from different combinations of the options available for each supply component and calculating the life-cycle costs of each supply strategy under a set of different scenarios reflecting the uncertainty of future events

  16. Exploration Drilling and Technology Demonstration At Fort Bliss

    Energy Technology Data Exchange (ETDEWEB)

    Barker, Ben; Moore, Joe [EGI; Segall, Marylin; Nash, Greg; Simmons, Stuart; Jones, Clay; Lear, Jon; Bennett, Carlon

    2014-02-26

    The Tularosa-Hueco basin in south-central New Mexico has long been known as an extensional area of high heat flow. Much of the basin is within the Fort Bliss military reservation, which is an exceptionally high value customer for power independent of the regional electric grid and for direct use energy in building climate control. A series of slim holes drilled in the 1990s established the existence of a thermal anomaly but not its practical value. This study began in 2009 with a demonstration of new exploration drilling technology. The subsequent phases reported here delivered a useful well, comparative exploration data sets and encouragement for further development. A production-size well, RMI56-5, was sited after extensive study of archival and newly collected data in 2010-2011. Most of 2012 was taken up with getting state and Federal authorities to agree on a lead agency for permitting purposes, getting a drilling permit and redesigning the drilling program to suit available equipment. In 2013 we drilled, logged and tested a 924 m well on the McGregor Range at Fort Bliss using a reverse circulation rig. Rig tests demonstrated commercial permeability and the well has a 7-inch slotted liner for use either in production or injection. An August 2013 survey of the completed well showed a temperature of 90 C with no reversal, the highest such temperature in the vicinity. The well’s proximity to demand suggests a potentially valuable resource for direct use heat and emergency power generation. The drilling produced cuttings of excellent size and quality. These were subjected to traditional analyses (thin sections, XRD) and to the QEMScan™ for comparison. QEMScan™ technology includes algorithms for determining such properties of rocks as density, mineralogy, heavy/light atoms, and porosity to be compared with direct measurements of the cuttings. In addition to a complete cuttings set, conventional and resistivity image logs were obtained in the open hole before

  17. Safety considerations concerning light water reactors in Sweden

    International Nuclear Information System (INIS)

    Nilsson, T.

    1977-01-01

    In 1975 the Swedish Nuclear Power Inspectorate was commissioned by the Government to perform a Reactor Safety Study concerning commercial light water reactors. The study will contain an account of: - rules and regulations for reactor designs; - operation experience of the Swedish nuclear power plants with international comparisons; - the development of reactor designs during the last 10 years; - demands and conditions for inspection and inspection methods; - nuclear power plant operation organization; - training of operators; and - the results of research into nuclear safety. The study is scheduled for completion by July 1st, 1977, however, this paper gives a summary of the results of the Reactor Safety Study already available. The paper contains detailed statistics concerning safety related occurrences and reactor scrams in Sweden from July 1st, 1974 until the beginning of 1977

  18. UAV Survey Data from Clifton Camp (ST56557330, Bristol, UK

    Directory of Open Access Journals (Sweden)

    Stephen Gray

    2015-05-01

    Full Text Available This data was collected via low-altitude UAV (Unmanned Aerial Vehicle survey of an area of Clifton Camp (ST565557330, best known for its Iron Age promontory fort. The dataset comprises of metadata records, near-vertical photographs and a derived 3D polygonal mesh. This dataset has been constructed with two kinds of reuse in mind: Firstly, the area surveyed is culturally rich and underexplored; while some of the non-natural features detected by this survey can be identified, others cannot. This data is intended to inform future investigations of the site. Secondly, the survey methodologies employed and the structuring of the resulting dataset are intended to act as an exemplar, a standard method of creating survey data while prioritising open technologies, and of organising UAV survey datasets to ensure maximum re-usability.

  19. Pulsed reactors: A dissenting view

    International Nuclear Information System (INIS)

    Ganev, I.Kh.; Orlov, V.V.

    1995-01-01

    The preceding article, by G.A. Ivanov et al., contains interesting estimates of the expanded production of plutonium in thermonuclear explosions initiated by plutonium charges. It must be noted that more than 40 years of efforts, despite some technical successes, have not led to a fast-reactor technology suitable for large-scale power production. This explains the incessant search for a nuclear technology for the future and the renewed interest in accelerator, hybrid, and explosive approaches to plutonium production. The success of such efforts will depend largely on the formulation of goals and the choice of the principal criteria. It is appropriate to discuss these issues here because the adoption of the rate of plutonium production or the plutonium doubling time as the principal criterion sets the stage for the repetition of previous errors. However, as a preliminary, I would like to question some categorical assertions that were made by Ivanov et al. without the presentation of adequate supporting data (the assertions that open-quotes the creation of an power industry on the basis of ordinary breeder reactors is practically impossibleclose quotes and that open-quotes adequate power generation in the 21st centuryclose quotes is impossible). In fact, it is simple to calculate that, given a realistic doubling time for fast reactors of ∼10 years and the plutonium produced by thermal reactors (around 10 12 W), it would be possible, if so desired, to introduce power far exceeding 10 14 W in the 21st century

  20. Slurry reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kuerten, H; Zehner, P [BASF A.G., Ludwigshafen am Rhein (Germany, F.R.)

    1979-08-01

    Slurry reactors are designed on the basis of empirical data and model investigations. It is as yet not possible to calculate the flow behavior of such reactors. The swarm of gas bubbles and cluster formations of solid particles and their interaction in industrial reactors are not known. These effects control to a large extent the gas hold-up, the gas-liquid interface and, similarly as in bubble columns, the back-mixing of liquids and solids. These hydrodynamic problems are illustrated in slurry reactors which constructionally may be bubble columns, stirred tanks or jet loop reactors. The expected effects are predicted by means of tests with model systems modified to represent the conditions in industrial hydrogenation reactors. In his book 'Mass Transfer in Heterogeneous Catalysis' (1970) Satterfield complained of the lack of knowledge about the design of slurry reactors and hence of the impossible task of the engineer who has to design a plant according to accepted rules. There have been no fundamental changes since then. This paper presents the problems facing the engineer in designing slurry reactors, and shows new development trends.

  1. Reactor safety

    International Nuclear Information System (INIS)

    Butz, H.P.; Heuser, F.W.; May, H.

    1985-01-01

    The paper comprises an introduction into nuclear physics bases, the safety concept generally speaking, safety devices of pwr type reactors, accident analysis, external influences, probabilistic safety assessment and risk studies. It further describes operational experience, licensing procedures under the Atomic Energy Law, research in reactor safety and the nuclear fuel cycle. (DG) [de

  2. Nuclear reactor

    International Nuclear Information System (INIS)

    Mysels, K.J.; Shenoy, A.S.

    1976-01-01

    A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

  3. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  4. RA Reactor

    International Nuclear Information System (INIS)

    1989-01-01

    This chapter includes the following: General description of the RA reactor, organization of work, responsibilities of leadership and operators team, regulations concerning operation and behaviour in the reactor building, regulations for performing experiments, regulations and instructions for inserting samples into experimental channels [sr

  5. Reactor physics

    International Nuclear Information System (INIS)

    Ait Abderrahim, H.

    1998-01-01

    Progress in research on reactor physics in 1997 at the Belgian Nuclear Research Centre SCK/CEN is described. Activities in the following four domains are discussed: core physics, ex-core neutron transport, experiments in Materials Testing Reactors, international benchmarks

  6. Fort Calhoun Station, Unit 2. License application, PSAR, general information

    International Nuclear Information System (INIS)

    1975-09-01

    Application for construction and operating licenses for Calhoun-2 Reactor is presented. Financial data concerning the Omaha Public Power District and the Nebraska Public Power District are included. (DCC)

  7. Reactor core

    International Nuclear Information System (INIS)

    Azekura, Kazuo; Kurihara, Kunitoshi.

    1992-01-01

    In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

  8. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

    A system for controlling combustible gases, it has been constituted at present such that the combustible gases are controlled by exhausting them to the wet well of a reactor container. In this system, however, there has been a problem, in a reactor container having plenums in addition to the wet well and the dry well, that the combustible gases in such plenums can not be controlled. In view of the above, in the present invention, suction ports or exhaust ports of the combustible gas control system are disposed to the wet well, the dry well and the plenums to control the combustible gases in the reactor container. Since this can control the combustible gases in the entire reactor container, the integrity of the reactor container can be ensured. (T.M.)

  9. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Tilliette, Z.

    1975-01-01

    A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

  11. 14. informal meeting on reactor noise

    International Nuclear Information System (INIS)

    1981-01-01

    The present booklet contains abstracts of papers from the 14th informal meeting on reactor noise held at St. Englmar in April 1981. The main topics dealt with are vibration and loose part monitoring, leak detection, noise theory and noise applications and in the final part data processing and pattern recognition techniques. (orig.)

  12. Electricity Generation from Geothermal Resources on the Fort Peck Reservation in Northeast Montana

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, Garry J. [Gradient Geophysics Inc., Missoula, MT (United States); Birkby, Jeff [Birkby Consulting LLC, Missoula, MT (United States)

    2015-05-12

    Tribal lands owned by Assiniboine and Sioux Tribes on the Fort Peck Indian Reservation, located in Northeastern Montana, overlie large volumes of deep, hot, saline water. Our study area included all the Fort Peck Reservation occupying roughly 1,456 sq miles. The geothermal water present in the Fort Peck Reservation is located in the western part of the Williston Basin in the Madison Group complex ranging in depths of 5500 to 7500 feet. Although no surface hot springs exist on the Reservation, water temperatures within oil wells that intercept these geothermal resources in the Madison Formation range from 150 to 278 degrees F.

  13. Anatomy of the Le Fort I segment: Are arterial variations a potential risk factor for avascular bone necrosis in Le Fort I osteotomies?

    Science.gov (United States)

    Bruneder, Simon; Wallner, Jürgen; Weiglein, Andreas; Kmečová, Ĺudmila; Egger, Jan; Pilsl, Ulrike; Zemann, Wolfgang

    2018-05-02

    Osteotomies of the Le Fort I segment are routine operations with low complication rates. Ischemic complications are rare, but can have severe consequences that may lead to avascular bone necrosis of the Le Fort I segment. Therefore the aim of this study was to investigate the blood supply and special arterial variants of the Le Fort I segment responsible for arterial hypoperfusion or ischemic avascular necrosis after surgery. The arterial anatomy of the Le Fort I segment's blood supply using 30 halved human cadaver head specimens was analyzed after complete dissection until the submicroscopic level. In all specimens the arterial variants of the Le Fort I segment and also the arterial diameters measured at two points were evaluated. The typical known vascularization pattern was apparent in 90% of all specimens, in which the ascending palatine (D1: 1,2 mm ± 0,34 mm; D2: 0,8 mm ± 0,34 mm) and ascending pharyngeal artery (D1: 1,3 mm ± 0,58 mm; D2: avascular segment necrosis after surgery. An individualized operation plan may prevent ischemic complications in at-risk patients. Copyright © 2018 European Association for Cranio-Maxillo-Facial Surgery. Published by Elsevier Ltd. All rights reserved.

  14. St. Anselm and Teaching

    Science.gov (United States)

    Evans, Gillian R.

    1976-01-01

    Discusses the educational principles and methods of the medieval philosopher and theologian, St. Anselm. Educational practices of medieval monastic schools, town schools, private tutors, and pastoral teaching are discussed. Available from: Taylor & Francis Ltd., P.O. Box 9137, Church Street Station, New York, N.Y. 10049. (Author/DB)

  15. Reactor building

    International Nuclear Information System (INIS)

    Maruyama, Toru; Murata, Ritsuko.

    1996-01-01

    In the present invention, a spent fuel storage pool of a BWR type reactor is formed at an upper portion and enlarged in the size to effectively utilize the space of the building. Namely, a reactor chamber enhouses reactor facilities including a reactor pressure vessel and a reactor container, and further, a spent fuel storage pool is formed thereabove. A second spent fuel storage pool is formed above the auxiliary reactor chamber at the periphery of the reactor chamber. The spent fuel storage pool and the second spent fuel storage pool are disposed in adjacent with each other. A wall between both of them is formed vertically movable. With such a constitution, the storage amount for spent fuels is increased thereby enabling to store the entire spent fuels generated during operation period of the plant. Further, since requirement of the storage for the spent fuels is increased stepwisely during periodical exchange operation, it can be used for other usage during the period when the enlarged portion is not used. (I.S.)

  16. Reactor container

    International Nuclear Information System (INIS)

    Shibata, Satoru; Kawashima, Hiroaki

    1984-01-01

    Purpose: To optimize the temperature distribution of the reactor container so as to moderate the thermal stress distribution on the reactor wall of LMFBR type reactor. Constitution: A good heat conductor (made of Al or Cu) is appended on the outer side of the reactor container wall from below the liquid level to the lower face of a deck plate. Further, heat insulators are disposed to the outside of the good heat conductor. Furthermore, a gas-cooling duct is circumferentially disposed at the contact portion between the good heat conductor and the deck plate around the reactor container. This enables to flow the cold heat from the liquid metal rapidly through the good heat conductor to the cooling duct and allows to maintain the temperature distribution on the reactor wall substantially linear even with the abrupt temperature change in the liquid metal. Further, by appending the good heat conductor covered with inactive metals not only on the outer side but also on the inside of the reactor wall to introduce the heat near the liquid level to the upper portion and escape the same to the cooling layer below the roof slab, the effect can be improved further. (Ikeda, J.)

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Rau, P.

    1980-01-01

    The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

  18. 2009 Federal Emergency Management Agency (FEMA) Topographic LiDAR: Fort Kent, Maine

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Camp Dresser McKee Inc. contracted with Sanborn Map Company to provide LiDAR mapping services for Fort Kent, Maine. Utilizing multi-return systems, Light Detection...

  19. Design Schematics for a Sustainable Parking Lot: Building 2-2332, ENRD Classroom, Fort Bragg, NC

    National Research Council Canada - National Science Library

    Stumpf, Annette

    2003-01-01

    ...) was tasked with planning a sustainable design "charrette" to explore and develop alternative parking lot designs that would meet Fort Bragg's parking needs, as well as its need to meet sustainable...

  20. Vendor Payments-Operation Mongoose, Fort Belvoir Defense Accounting Office and Rome Operating Location

    National Research Council Canada - National Science Library

    Lane, F

    1996-01-01

    .... Due to the impending closure of the Defense Accounting Office at Fort Belvoir and the anticipated consolidation to the Rome Operating Location, New York, we did not perform a review of the management...

  1. Final Sampling and Analysis Plan for Background Sampling, Fort Sheridan, Illinois

    National Research Council Canada - National Science Library

    1995-01-01

    .... This Background Sampling and Analysis Plan (BSAP) is designed to address this issue through the collection of additional background samples at Fort Sheridan to support the statistical analysis and the Baseline Risk Assessment (BRA...

  2. Freight Advanced Traveler Information System (FRATIS) - Dallas-Fort Worth (DFW) prototype : final report.

    Science.gov (United States)

    This is the Final Report for the FRATIS Dallas-Fort Worth DFW prototype system. The FRATIS prototype in : DFW consisted of the following components: optimization algorithm, terminal wait time, route specific : navigation/traffic/weather, and advanced...

  3. Notification: Hotline Complaint – Drinking Water Treatment Plant at the Fort Belknap Indian Community

    Science.gov (United States)

    Project #OA-FY13-0076, November 13, 2012. On March 22, 2012, the Office of Inspector General (OIG) received a hotline complaint on the construction of the Drinking Water Treatment Plant (DWTP) at the Fort Belknap Indian Community.

  4. Biological Assessment of Streams Associated with the Northern Training Complex at Fort knox, Kentucky, August 2000

    National Research Council Canada - National Science Library

    Payne, Berry

    2001-01-01

    .... The benthic macroinvertebrate aspect of the U.S. Environmental Protection Agency's Rapid Bioassessment Protocol was applied in August 2000 to selected streams likely to be affected by proposed improvements of training facilities on Fort Knox...

  5. 75 FR 24930 - Fort Bliss (Texas) Army Growth and Force Structure Realignment Final Environmental Impact...

    Science.gov (United States)

    2010-05-06

    ...-PWE, Building 624, Taylor Road, Fort Bliss, TX 79916-6812; e- mail: [email protected] . FOR... Regional Branch Library, 551 Redd Road. In Las Cruces (NM), the New Mexico State University Zuhl Library...

  6. Feasibility Study for an Off-Post, Primary Care Clinic at Fort Campbell, Kentucky

    National Research Council Canada - National Science Library

    Kvalevog, Kristen J

    2005-01-01

    .... Over 90,679 beneficiaries currently live in -the-Fort Campbell-catchment area and receive primary care at Blanchfield Army Community Hospital through the Red, White, Blue, Gold, and Young Eagle Clinics...

  7. Identifying 21st Century Capabilities

    Science.gov (United States)

    Stevens, Robert

    2012-01-01

    What are the capabilities necessary to meet 21st century challenges? Much of the literature on 21st century skills focuses on skills necessary to meet those challenges associated with future work in a globalised world. The result is a limited characterisation of those capabilities necessary to address 21st century social, health and particularly…

  8. Site Investigations with the Site Characterization and Analysis Penetrator System at Fort Dix, New Jersey

    Science.gov (United States)

    1993-07-01

    rod system or through a tremie tube ; both procedures were used interchangeably at Fort Dix to demon- strate the efficiency and effectiveness of each...allows delivery through either a l/4-in.-diam grout tube or a 3/8-in.-diam rout tube . The grout used at Fort Dix consisted of a mixture of water and... microfine , blended Portland cement (Lehigh Geocem’, Leeds, Alabama). The grout is a suspension of a uniformly produced cement clinker interground with

  9. Cannon Fire Soon to Accompany Bugle Call at Fort Detrick | Poster

    Science.gov (United States)

    Beginning June 14, the familiar bugle calls at Fort Detrick will be joined by a special percussion instrument: a cannon. A single cannon shot will be fired on the first note of “Reveille,” which signals the start of each day and is accompanied by the raising of the American flag. “Reveille” sounds at 6:30 a.m. At 5 p.m., Fort Detrick plays “Retreat,” which alerts the post that

  10. Inpatient Behavioral Health Recapture A Busiess Case Analysis at Evans Army Community Hospital Fort Carson, Colorado

    Science.gov (United States)

    2009-07-20

    and Obstetrics /Gynecology. Inpatient care includes Obstetrics , Intensive Care, and Post Anesthesia Care/Same Day Surgery. EACH Mission: Delivering...charged with murder in Iraq shooting deaths, 2009). EACH Inpt Psych 13 Fort Carson has not been immune to the increase in suicides and violence among...to identify Soldiers with PTSD symptoms. In 2008, however, attention returned to Fort Carson as a number of local homicides and other violence tied

  11. How Do Le Fort-Type Fractures Present in a Pediatric Cohort?

    Science.gov (United States)

    Macmillan, Alexandra; Lopez, Joseph; Luck, J D; Faateh, Muhammad; Manson, Paul; Dorafshar, Amir H

    2018-05-01

    Le Fort-type fractures are very rare in children, and there is a paucity of literature presenting their frequency and characteristics. The purpose of this study was to determine the etiology, frequency, and fracture patterns of children with severe facial trauma associated with pterygoid plate fractures in a pediatric cohort. We performed a retrospective cohort study of all children aged younger than 16 years with pterygoid plate and facial fractures who presented to our institute between 1990 and 2010. Patient charts and radiologic records were reviewed for demographic and fracture characteristics. Patients were categorized into 2 groups as per facial fracture pattern: non-Le Fort-type fractures (group A) and Le Fort-type fractures (group B). Other variables including dentition age, frontal sinus development, mechanism of injury, injury severity, and concomitant injuries were recorded. Univariate methods were used to compare groups. We identified 24 children; 25% were girls, and 20.8% were of nonwhite race. Most presented with Le Fort-type fracture patterns (group B, 66.7%). Age was significantly different between group A and group B (mean, 5.9 years and 9.9 years, respectively; P = .009). No significant differences in Injury Severity Score, rate of operative repair, and length of stay were found between groups. Most children with severe facial fractures and pterygoid plate fractures presented with Le Fort-type fracture patterns in our cohort. The mean age of children with Le Fort-type fractures was greater than in those with non-Le Fort-type patterns. However, Le Fort-type fractures did occur in younger children with deciduous and mixed dentition. Copyright © 2017 American Association of Oral and Maxillofacial Surgeons. Published by Elsevier Inc. All rights reserved.

  12. Rumlige fortællinger fra mobilt og web-baseret GIS

    DEFF Research Database (Denmark)

    Møller-Jensen, Lasse

    2009-01-01

    Denne artikel handler om begrebet rumlige fortællinger med anvendelse af fortællingshenvisninger, og disses potentielle rolle ved implementation af fleksible og tematiske turistinformationssystemer. Artiklen fokuserer på brugen af mobile, positionsbekendte enheder, såsom visse PDA'er og smartphon......, samt på web-gis. Der præsenteres to anvendelseseksempler: et fra det centrale København og et fra et område nær Accra, Ghana....

  13. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.

    1976-01-01

    A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Miyashita, Akio.

    1981-01-01

    Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

  15. Reactor container

    International Nuclear Information System (INIS)

    Abe, Yoshihito; Sano, Tamotsu; Ueda, Sabuo; Tanaka, Kazuhisa.

    1987-01-01

    Purpose: To improve the liquid surface disturbance in LMFBR type reactors. Constitution: A horizontal flow suppressing mechanism mainly comprising vertical members is suspended near the free liquid surface of coolants in the upper plenum. The horizontal flow of coolants near the free liquid surface is reduced by the suppressing mechanism to effectively reduce the surface disturbance. The reduction in the liquid surface disturbance further prevails to the entire surface region with no particular vertical variations to the free liquid surface to remarkably improve the preventive performance for the liquid surface disturbance. Accordingly, it is also possible to attain the advantageous effects such as prevention for the thermal fatigue in reactor vessel walls, reactor upper mechanisms, etc. and prevention of burning damage to the reactor core due to the reduction of envolved Ar gas. (Kamimura, M.)

  16. 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.

  17. NUCLEAR REACTOR

    Science.gov (United States)

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  18. Breeder reactors

    International Nuclear Information System (INIS)

    Gollion, H.

    1977-01-01

    The reasons for the development of fast reactors are briefly reviewed (a propitious neutron balance oriented towards a maximum uranium burnup) and its special requirements (cooling, fissile material density and reprocessing) discussed. The three stages in the French program of fast reactor development are outlined with Rapsodie at Cadarache, Phenix at Marcoule, and Super Phenix at Creys-Malville. The more specific features of the program of research and development are emphasized: kinetics and the core, the fuel and the components [fr

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Schulze, I.; Gutscher, E.

    1980-01-01

    The core contains a critical mass of UN or U 2 N 3 in the form of a noncritical solution with melted Sn being kept below a N atmosphere. The lining of the reactor core consists of graphite. If fission progresses part of the melted metal solution is removed and cleaned from fission products. The reactor temperatures lie in the range of 300 to 2000 0 C. (Examples and tables). (RW) [de

  20. Reactor technology

    International Nuclear Information System (INIS)

    Erdoes, P.

    1977-01-01

    This is one of a series of articles discussing aspects of nuclear engineering ranging from a survey of various reactor types for static and mobile use to mention of atomic thermo-electric batteries of atomic thermo-electric batteries for cardiac pacemakers. Various statistics are presented on power generation in Europe and U.S.A. and economics are discussed in some detail. Molten salt reactors and research machines are also described. (G.M.E.)

  1. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  2. Fast reactors

    International Nuclear Information System (INIS)

    Vasile, A.

    2001-01-01

    Fast reactors have capacities to spare uranium natural resources by their breeding property and to propose solutions to the management of radioactive wastes by limiting the inventory of heavy nuclei. This article highlights the role that fast reactors could play for reducing the radiotoxicity of wastes. The conversion of 238 U into 239 Pu by neutron capture is more efficient in fast reactors than in light water reactors. In fast reactors multi-recycling of U + Pu leads to fissioning up to 95% of the initial fuel ( 238 U + 235 U). 2 strategies have been studied to burn actinides: - the multi-recycling of heavy nuclei is made inside the fuel element (homogeneous option); - the unique recycling is made in special irradiation targets placed inside the core or at its surroundings (heterogeneous option). Simulations have shown that, for the same amount of energy produced (400 TWhe), the mass of transuranium elements (Pu + Np + Am + Cm) sent to waste disposal is 60,9 Kg in the homogeneous option and 204.4 Kg in the heterogeneous option. Experimental programs are carried out in Phenix and BOR60 reactors in order to study the feasibility of such strategies. (A.C.)

  3. Oliver St John Gogarty.

    Science.gov (United States)

    Clarke, R W

    1997-01-01

    Oliver St John Gogarty--Otolaryngologist to fashionable Edwardian Dublin--was a distinguished poet and a Senator in the fledgling Irish Free State after its establishment in 1922. He numbered amongst his acquaintances the poet William Butler Yeats, the novelist James Joyce and a host of political and literary persona who helped to shape modern Ireland. He was satirised as 'stately plump Buck Mulligan' in Joyce's novel Ulysses.

  4. Habitat Restoration/Enhancement Fort Hall Reservation : 2008 Annual Report.

    Energy Technology Data Exchange (ETDEWEB)

    Osborne, Hunter [Shoshone Bannock Tribes

    2009-07-23

    Habitat enhancement, protection and monitoring were the focus of the Resident Fisheries Program during 2008. Enhancement and protection included sloping, fencing and planting wetlands plugs at sites on Spring Creek (Head-waters). Many previously constructed instream structures (rock barbs and wing dams) were repaired throughout the Fort Hall Indian Reservation (Reservation). Physical sampling during 2008 included sediment and depth measurements (SADMS) in Spring Creek at the Car Removal site. SADMS, used to track changes in channel morphology and specifically track movements of silt through Bottoms stream systems were completed for 5 strata on Spring Creek. Water temperature and chemistry were monitored monthly on Spring Creek, Clear Creek, Diggie Creek, and Portneuf (Jimmy Drinks) and Blackfoot rivers. Fish population densities and biomass were sampled in five reservation streams which included nine sites. Sampling protocols were identical to methods used in past years. Numbers of fish in Spring Creek series remained relatively low, however, there was an increase of biomass overall since 1993. Salmonid fry densities were monitored near Broncho Bridge and were similar to 2006, and 2007, however, as in years past, high densities of macrophytes make it very difficult to see fry in addition to lack of field technicians. Mean catch rate by anglers on Bottoms streams stayed the same as 2007 at 1.5/hr. Numbers of fish larger than 18-inches caught by anglers increased from 2007 at .20 to .26/hr.

  5. Fate of TCE in heated Fort Lewis soil.

    Science.gov (United States)

    Costanza, Jed; Fletcher, Kelly E; Löffler, Frank E; Pennell, Kurt D

    2009-02-01

    This study explores the transformation of trichloroethene (TCE) caused by heating contaminated soil and groundwater samples obtained from the East Gate Disposal Yard (EGDY) located in Fort Lewis, WA. After field samples transferring into glass ampules and introducing 1.5 micromol of TCE, the sealed ampules were incubated at temperatures of 25, 50, and 95 degrees C for periods of up to 95.5 days. Although TCE was completely transformed into cis-1,2-dichloroethene (cis-DCE) after 42 days at 25 degrees C by microbial activity, this transformation was not observed at 50 or 95 degrees C. Chloride levels increased after 42 days at 25 degrees C corresponding to the mass of TCE transformed to cis-DCE, were constant at 50 degrees C, and increased at 95 degrees C yielding a TCE degradation half-life of 1.6-1.9 years. These findings indicate that indigenous microbes contribute to the partial dechlorination of TCE to cis-DCE at temperatures of less than 50 degrees C, whereas interphase mass transfer and physical recovery of TCE will predominate over in situ degradation processes at temperatures of greater than 50 degrees C during thermal treatment at the EGDY site.

  6. An overview of the 2009 Fort Hood Robotics Rodeo

    Science.gov (United States)

    Norberg, Seth

    2010-04-01

    The Robotics Rodeo held from 31 August to 3 September 2009 at Fort Hood, Texas, had three stated goals: educate key decision makers and align the robotics industry; educate Soldiers and developers; and perform a live market survey of the current state of technologies to encourage the development of robotic systems to support operational needs. Both events that comprised the Robotics Rodeo, the Extravaganza and the robotic technology observation, demonstration and discussion (RTOD2) addressed these stated goals. The Extravaganza was designed to foster interaction between the vendors and the visitors who included the media, Soldiers, others in the robotics industry and key decision makers. The RTOD2 allowed the vendors a more private and focused interaction with the subject matter experts teams, this was the forum for the vendors to demonstrate their robotic systems that supported the III Corps operational needs statements that are focused on route clearance, convoy operations, persistent stare, and robotic wingman. While the goals of the Rodeo were achieved, the underlying success from the event is the development of a new business model that is focused on collapsing the current model to get technologies into the hands of our warfighters quicker. This new model takes the real time data collection from the Rodeo, the Warfighter Needs from TRADOC, the emerging requirements from our current engagements, and assistance from industry partners to develop a future Army strategy for the rapid fielding of unmanned systems technologies.

  7. Soil classification using CPTu in Fort McMurray

    Energy Technology Data Exchange (ETDEWEB)

    Elbanna, M. [AMEC Earth and Environmental, Nanaimo, BC (Canada); El Sabbagh, M. [AMEC Earth and Environmental, Burnaby, BC (Canada); Sharp, J. [ConeTec Investigations Ltd., Richmond, BC (Canada)

    2009-07-01

    This paper evaluated 4 piezocone penetration testing (CPTu) classification methods using data from 3 different sites near Fort McMurray in northern Alberta. For comparative purposes, other in-situ tests, field observations, and laboratory tests were performed at all sites in close proximity to the CPTu soundings. The study evaluated pleistocene sand and sand till deposits with low fines content. Profiling these deposits is necessary because they are often used as filler material for earth retaining structures in many oilsands projects. The study also evaluated pleistocene clay and clay tills that are often used as low permeability material for seepage control. In thick layers, pleistocene clay is known to cause foundation problems. CPTu with dissipation data was shown to be a useful tool in geotechnical engineering practice to provide near continuous soil profiling and material properties. CPTu tip resistance and sleeve friction combined with pore pressure measurement provided useful evaluation of subsurface soil types. It was concluded that although all of the CPTu classification charts provided reasonable soil classification in typical soil conditions, local experience and understanding of soil behaviour is needed to make an appropriate selection of the most applicable charts in a given geological condition. 7 refs., 11 figs.

  8. Permafrost delineation for remediation planning : Fort Wainwright, Alaska

    Energy Technology Data Exchange (ETDEWEB)

    Astley, B. [Cold Regions Research and Engineering Laboratory, Anchorage, AK (United States); Snyder, C. [YEC Inc., Valley Cottage, NJ (United States); Delaney, A. [Cold Regions Research and Engineering Laboratory, Fairbanks, AK (United States); Arcone, S.; Lawson, D. [Cold Regions Research and Engineering Laboratory, Hanover, NH (United States)

    2003-07-01

    In the summer of 1999, geophysical and hydrogeological surveys were conducted at the Birch Hill Tank Farm and Truck Fill Stand in Fort Wainwright, Alaska to assess the distribution of benzene, 1,2-dichloroethane, and 1,2-dibromoethane. The Birch Hill site consists of a silt, sand and gravel fluvial deposit that overlies bedrock. Permafrost occurs discontinuously throughout the alluvium and underlying bedrock, resulting in a complex aquifer distribution. The bedrock beneath the Tank Farm is highly fractured and faulted with a weathered horizon that is 30 meters thick. The goal of this study was to map the discontinuous permafrost and aquifers in the alluvial deposits and weathered bedrock zone for the purpose of delineating bedrock depth and structural features that influence ground water flow. Several methods were used to define subsurface conditions, including borehole logs, DC resistivity, and ground-penetrating radar. A 3-D hydrogeologic model was used to develop a ground water flow model used to determine contaminant migration pathways and rates. The permafrost configuration was found to be the most important boundary condition in this model. 7 refs., 1 tab., 5 figs.

  9. Environmental Assessment Addressing the 301st Fighter Wing Managed Airspace, Naval Air Station Joint Reserve Base, Fort Worth, Texas

    Science.gov (United States)

    2009-05-01

    Salle County McMullen County Terrell County Blanco County Comal County Texas Oklahoma Dallas- Ft. Worth San Antonio Austin Tulsa Oklahoma City Abilene...County Young County Crockett County Glasscock County Irion County Midland County Brown County Reagan County Sterling County Terrell County Upton...7909 Karl May Drive Waco, TX 76708 Margaret Wood Brown County Clerk 200 South Broadway Brownwood, TX 76801 Jo Ann Hale Coleman County

  10. 78 FR 114 - Notice of January 23 and 24, 2013 Meeting for Fort Hancock 21st Century Advisory Committee

    Science.gov (United States)

    2013-01-02

    ... Island, NY 10305, at (718) 354-4710 or email: admin@forthancock21stcentury.org , or visit the Advisory Committee Web site at www.forthancock21stcentury.org . SUPPLEMENTARY INFORMATION: Under section 10(a)(2) of...

  11. 77 FR 20046 - Establishment of the Gateway National Recreation Area Fort Hancock 21st Century Advisory Committee

    Science.gov (United States)

    2012-04-03

    ..., Interior. ACTION: Notice and Call for Nominations. SUMMARY: The Secretary of the Interior (Secretary) is... resume listing their name, title, address, telephone, email, and fax number. DATES: Written nominations...

  12. Next Generation Reactors in Korea

    International Nuclear Information System (INIS)

    Oh, Yongshick; Choi, Youngsang; Park, Keecheol

    1990-01-01

    In Korea, nuclear power will be continuously needed to meet the trend of steady increase in electricity demand. But in relation to the further development of nuclear energy, there are still many uncertainties to be solved such as power demand forecast, site availability, thermal energy utilization and technology enhancement for economic and safety. To cope with those uncertainties effectively and to proceed the nuclear projects uninterruptedly, KEPCO decided to initiate two research project. i. e., one is 'the outlook and developmental strategy of nuclear energy for the early 21st century in the R. O. K' and the other is 'the feasibility study on the advanced reactors in Korea. Prospects of nuclear energy in Korea was overviewed and recommendations from the industry were introduced. It is strong opinion of Korea nuclear industry that nuclear policy should be changed from the support policy to the target management policy. In the point of reactor strategy, the life of light water reactor technology might be longer than expected before in Korea and it is emphasized that good maintenance of light water reactor technology and smooth transition program to the advanced technologies should be carefully considered. There are differences in the opinions between preferences to the evolutionary and/or passive, inherently safe reactors but, in the long-term point of view, it is judged to be desirable to have alternatives

  13. ST Safety What's on?

    CERN Document Server

    Jacot, C

    2001-01-01

    Du premier coup de pioche jusqu'à la mise en place de la tente abritant le buffet offert lors de la cérémonie d'achèvement d'un projet, la division ST est sur les rangs et apporte son savoir faire. Ces projets nombreux et pluridisciplinaires sont étudiés, se réalisent, s'exploitent et font l'objet de maintenance jusqu'à leur démantèlement. Durant leurs cycles de vie, la division ST avance les compétences professionnelles spécifiques à chaque corps de métiers incluant le recours à de nombreux contrats avec des entreprises extérieures pour réaliser les travaux sur les sites du CERN. Pour en conserver la maîtrise, la division ST applique une gestion de projet, une assurance de la qualité et s'attache tout particulièrement à la sécurité, à la réglementation, à la prévention des risques et la gestion de ceux-ci. Sous la supervision du DSO, un ensemble de procédures et de démarches a été développé afin de faciliter les tâches du maître d'ouvrage et des entreprises contractantes, c...

  14. Roegneria alashanica Keng: a species with the StStStYStY genome constitution.

    Science.gov (United States)

    Wang, Richard R-C; Jensen, Kevin B

    2017-06-01

    The genome constitution of tetraploid Roegneria alashanica Keng has been in question for a long time. Most scientific studies have suggested that R. alashanica had two versions of the St genome, St 1 St 2 , similar to that of Pseudoroegneria elytrigioides (C. Yen & J.L. Yang) B.R. Lu. A study, however, concluded that R. alashanica had the StY genome formula typical for tetraploid species of Roegneria. For the present study, R. alashanica, Elymus longearistatus (Bioss.) Tzvelev (StY genomes), Pseudoroegneria strigosa (M. Bieb.) Á. Löve (St), Pseudoroegneria libanoctica (Hackel) D.R. Dewey (St), and Pseudoroegneria spicata (Pursh) Á. Löve (St) were screened for the Y-genome specific marker B14(F+R) 269 . All E. longearistatus plants expressed intense bands specific to the Y genome. Only 6 of 10 R. alashanica plants exhibited relatively faint bands for the STS marker. Previously, the genome in species of Pseudoroegneria exhibiting such faint Y-genome specific marker was designated as St Y . Based on these results, R. alashanica lacks the Y genome in E. longearistatus but likely possess two remotely related St genomes, St and St Y . According to its genome constitution, R. alashanica should be classified in the genus Pseudoroenera and given the new name Pseudoroegneria alashanica (Keng) R.R.-C. Wang and K.B. Jensen.

  15. Generation IV reactors: reactor concepts

    International Nuclear Information System (INIS)

    Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

    2003-01-01

    Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

  16. Research reactors - an overview

    International Nuclear Information System (INIS)

    West, C.D.

    1997-01-01

    A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs

  17. Comparison of maxillary stability after Le Fort I osteotomy for occlusal cant correction surgery and maxillary advanced surgery.

    Science.gov (United States)

    Ueki, Koichiro; Hashiba, Yukari; Marukawa, Kohei; Yoshida, Kan; Shimizu, Chika; Nakagawa, Kiyomasa; Yamamoto, Etsuhide

    2007-07-01

    To compare postoperative maxillary stability following Le Fort I osteotomy for the correction of occlusal cant as compared with conventional Le Fort I osteotomy for maxillary advancement. The subjects were 40 Japanese adults with jaw deformities. Of these, 20 underwent a Le Fort I osteotomy and intraoral vertical ramus osteotomy (IVRO) to correct asymmetric skeletal morphology and inclined occlusal cant. The other 20 patients underwent a Le Fort I osteotomy and sagittal split ramus osteotomy (SSRO) to advance the maxilla. Lateral and posteroanterior cephalograms were taken postoperatively and assessed statistically. Thereafter, the 2 groups were followed for time-course changes. There was no significant difference between the 2 groups with regard to time-course changes during the immediate postoperative period. This suggests that maxillary stability after Le Fort I osteotomy for cant correction does not differ from that after Le Fort I osteotomy for maxillary advancement.

  18. OECD Halden reactor project

    International Nuclear Information System (INIS)

    1977-01-01

    The activities of the OECD Halden Reactor Project for the year 1975 are summarized. The period under review is the last year of the three year joint programme which commenced on 1st January, 1973. The main items reported upon are: process supervision and control, test fuel irradiation and fuel research, reactor operations, and administration and finance. The process supervision and control work has been concentrated in two fields: methods development for core surveillance and control, and systems development for operator-process communication. As for fuel test, investigations of the densification phenomenon have continued through irradiations to a maximum of about 16000MWd/tUO 2 . Axial and radial deformations of fuel rods are studied, with the effect of power transients upon the dimensional stability of fuel rods, and fuel-cladding heat transfer and fuel temperature. Thermal models for steady state and transient heat transfer in fuel rods have been developed and the work on thermomechanical models of claddings shows considerable promise

  19. The Fall of Fort Eben Emael: The Effects of Emerging Technologies on the Successful Completion of Military Objectives

    Science.gov (United States)

    2004-06-18

    of Sickle,” World War II Magazine, November 2003, 59. 11Ibid., 60. 12Abbeville is 100 miles north of Paris near the English Channel. 13T. N. Mout... catacombs of Fort Eben Emael. A further understanding of the dynamics of the fort and her defenders can be gained by the knowledge that that fort was...Green Devils German Paratroopers 1939-45 ( Paris , France: Histories & Collections, 1997), 27. Helmut Wenzl (Left in photo) Born - 10 March

  20. Proceedings of the European Research Reactor Conference - RRFM 2013 Transactions

    International Nuclear Information System (INIS)

    2013-01-01

    In 2013 RRFM, the European Research Reactor Conference is jointly organised by ENS and Atomexpo LLC. This time the Research Reactor community meet in St. Petersburg, Russia. The conference programme will revolve around a series of Plenary Sessions dedicated to the latest global developments with regards to research reactor technology and management. Parallel sessions will focus on all areas of the Fuel Cycle of Research Reactors, their Utilisation, Operation and Management as well as specific research projects and innovative methods in research reactor analysis and design. In 2013 the European Research Reactor Conference will for the first time give special attention to complementary safety assessments of Research Reactors, following the Fukushima-Dai-Ichi NPP's Accident. (authors)

  1. Low footwall accelerations and variable surface rupture behavior on the Fort Sage Mountains fault, northeast California

    Science.gov (United States)

    Briggs, Richard W.; Wesnousky, Steven G.; Brune, James N.; Purvance, Matthew D.; Mahan, Shannon

    2013-01-01

    The Fort Sage Mountains fault zone is a normal fault in the Walker Lane of the western Basin and Range that produced a small surface rupture (L 5.6 earthquake in 1950. We investigate the paleoseismic history of the Fort Sage fault and find evidence for two paleoearthquakes with surface displacements much larger than those observed in 1950. Rupture of the Fort Sage fault ∼5.6  ka resulted in surface displacements of at least 0.8–1.5 m, implying earthquake moment magnitudes (Mw) of 6.7–7.1. An older rupture at ∼20.5  ka displaced the ground at least 1.5 m, implying an earthquake of Mw 6.8–7.1. A field of precariously balanced rocks (PBRs) is located less than 1 km from the surface‐rupture trace of this Holocene‐active normal fault. Ground‐motion prediction equations (GMPEs) predict peak ground accelerations (PGAs) of 0.2–0.3g for the 1950 rupture and 0.3–0.5g for the ∼5.6  ka paleoearthquake one kilometer from the fault‐surface trace, yet field tests indicate that the Fort Sage PBRs will be toppled by PGAs between 0.1–0.3g. We discuss the paleoseismic history of the Fort Sage fault in the context of the nearby PBRs, GMPEs, and probabilistic seismic hazard maps for extensional regimes. If the Fort Sage PBRs are older than the mid‐Holocene rupture on the Fort Sage fault zone, this implies that current GMPEs may overestimate near‐fault footwall ground motions at this site.

  2. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Prescott, R F; George, B V; Baglin, C J

    1978-05-10

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1978-01-01

    Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

  4. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  5. Lip-nasal aesthetics following Le Fort I osteotomy.

    Science.gov (United States)

    Rosen, H M

    1988-02-01

    Forty-one patients undergoing Le Fort I osteotomy for superior and/or anterior repositioning of the maxilla were prospectively studied for changes in soft-tissue morphology of the nasomaxillary region. Nasal parameters studied were changes in interalar rim width and nasal tip projection. It was observed that alar rim width increases with anterior and/or superior repositioning of the maxilla, but increases in nasal tip projection occur only when there is an anterior vector of maxillary movement. These nasal changes could not be quantitatively correlated to magnitude of maxillary movement. Lip changes studied were the horizontal displacement at the vermilion border and subnasale versus that of the incisal edge and point A, respectively, when the maxilla is sagittally advanced and the vertical shortening of the lip versus that of the incisal edge when the maxilla is shortened. Using linear regression analysis, horizontal displacement of the upper lip at the vermilion border was 0.82 +/- 0.13 mm for every 1 mm of maxillary advancement at the incisal edge (p less than 0.001) and 0.51 +/- 0.13 at the subnasale for every 1 mm of maxillary advancement at point A (p less than 0.001). Eighty percent of patients undergoing maxillary intrusive procedures had lip shortening ranging from 20 to 50 percent of the vertical maxillary reduction. Surprisingly, no statistically significant correlation could be demonstrated for lip shortening versus extent of vertical maxillary reduction. Previous literature in disagreement with these findings is discussed. Guidelines for treatment planning utilizing these data are suggested.

  6. Slip Potential of Faults in the Fort Worth Basin

    Science.gov (United States)

    Hennings, P.; Osmond, J.; Lund Snee, J. E.; Zoback, M. D.

    2017-12-01

    Similar to other areas of the southcentral United States, the Fort Worth Basin of NE Texas has experienced an increase in the rate of seismicity which has been attributed to injection of waste water in deep saline aquifers. To assess the hazard of induced seismicity in the basin we have integrated new data on location and character of previously known and unknown faults, stress state, and pore pressure to produce an assessment of fault slip potential which can be used to investigate prior and ongoing earthquake sequences and for development of mitigation strategies. We have assembled data on faults in the basin from published sources, 2D and 3D seismic data, and interpretations provided from petroleum operators to yield a 3D fault model with 292 faults ranging in strike-length from 116 to 0.4 km. The faults have mostly normal geometries, all cut the disposal intervals, and most are presumed to cut into the underlying crystalline and metamorphic basement. Analysis of outcrops along the SW flank of the basin assist with geometric characterization of the fault systems. The interpretation of stress state comes from integration of wellbore image and sonic data, reservoir stimulation data, and earthquake focal mechanisms. The orientation of SHmax is generally uniform across the basin but stress style changes from being more strike-slip in the NE part of the basin to normal faulting in the SW part. Estimates of pore pressure come from a basin-scale hydrogeologic model as history-matched to injection test data. With these deterministic inputs and appropriate ranges of uncertainty we assess the conditional probability that faults in our 3D model might slip via Mohr-Coulomb reactivation in response to increases in injected-related pore pressure. A key component of the analysis is constraining the uncertainties associated with each of the principal parameters. Many of the faults in the model are interpreted to be critically-stressed within reasonable ranges of uncertainty.

  7. Spacer st4ructure

    International Nuclear Information System (INIS)

    Masetti, W.R.

    1978-01-01

    A spacer structure is described for maintaining a spaced relation between a plurality of generally parallel fuel rods within a housing in a nuclear reactor. The spacer structure is comprised of a grid pattern of ribs slotted to interlock with each other. The slots are arranged in such a way that when the ribs are welded to each other, the weld shrinkage is distributed uniformly in all directions to reduce or eliminate the amount of rework necessary in manufacturing the spacer structure

  8. Nuclear reactor

    International Nuclear Information System (INIS)

    Scholz, M.

    1976-01-01

    An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

  9. NEUTRONIC REACTOR

    Science.gov (United States)

    Wigner, E.P.; Weinberg, A.W.; Young, G.J.

    1958-04-15

    A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

  10. Separation of silicon carbide-coated fertile and fissile particles by gas classification

    International Nuclear Information System (INIS)

    Vaughen, V.C.A.

    1976-07-01

    The separation of 235 U and 233 U in the reprocessing of HTGR fuels is a key feature of the feed-breed fuel cycle concept. This is attained in the Fort St. Vrain (FSV) reactor by coating the fissile (Th- 235 U) particles and the fertile (Th- 233 U) particles separately with silicon carbide (SiC) layers to contain the fission products and to protect the kernels from burning in the head-end reprocessing steps. Pneumatic (gas) classification based on size and density differences is the reference process for separating the SiC-coated particles into fissile and fertile streams for subsequent handling. Terminal velocities have been calculated for the +- 2 sigma ranges of particle sizes and densities for ''Fissile B''--''Fertile A'' particles used in the FSV reactor. Because of overlapping particle fractions, a continuous pneumatic separator appears infeasible; however, a batch separation process can be envisioned. Changing the gas from air to CO 2 and/or the temperature to 300 0 C results in less than 10 percent change in calculated terminal velocities. Recently reported work in gas classification is discussed in light of the theoretical calculations. The pneumatic separation of fissile and fertile particles needs more study, specifically with regard to (1) measuring the recoveries and separation efficiencies of actual fissile and fertile fractions in the tests of the pneumatic classifiers; and (2) improving the contactor design or flowsheet to avoid apparent flow separation or flooding problems at the feed point when using the feed rates required for the pilot plant

  11. Preliminary draft: environmental impact statement for Hot Engineering Test Project at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Boyle, J.W.; Baxter, B.J.; Carpenter, J.A.

    1978-08-01

    The project considered is the Hot Engineering Test Project (HETP), which is to be located in largely existing facilities at Oak Ridge National Laboratory (ORNL). The project is a part of the National High Temperature Gas-Cooled Reactor Fuel Recycle Program, which seeks to demonstrate the technological feasibility of the recycle processes. The HETP will attempt to confirm the operability of the processes (proven feasible in cold or nonradioactive, benchtop experimentation) under the more realistic radioactive condition. As such, the operation will involve the reprocessing and refabrication of spent HTGR fuel rods obtained from the Fort St. Vrain reactor. The reference fuel is highly enriched uranium. No significant radiological impacts are expected from routine operation of the facility to any biota or ecosystem. Concentrations of one or more radionuclides in Whiteoak Lake will increase as a result of the combination of HETP wastes with other ORNL wastes. Nonradiological effects from construction activities and routine operation should be insignificant on land and water use and on terrestrial and aquatic ecosystems. No significant socioeconomic impacts should occur from either construction or operation of the facility. Some conservative accident scenarios depict significant releases of radioactivity. Effects should be localized and would not be severe for all but the most unlikely of such incidents. No significant long-term commitment of resources is expected to be required for the project. Nor are any large quantities of scarce or critical resources likely to be irreversibly or irretrievably committed to the project. Principal alternatives considered were: relocation of the project site, postponement of the project schedule, project cancellation, and chemical process variations

  12. Spherical torus (ST) concept and its reactor implications

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Lazarus, E.A.; Miller, R.L.; Carreras, B.A.; Hogan, J.T.; Krakowski, R.A.; Seed, T.J.; Zubrin, R.M.; Schnurr, N.M.

    1986-01-01

    A brief description of the spherical torus design is given. The design concept includes resistive demountable toroidal field coils, poloidal divertor for impurity control, oscillating-field current maintenance, RF initiation and ramp-up of the plasma current, and flowing liquid-metal breeding blanket. 4 refs., 6 figs

  13. ifo Konjunkturprognose 2015/2017: Verhaltener Aufschwung setzt sich fort

    OpenAIRE

    Wollmershäuser, Timo; Nierhaus, Wolfgang; Berg, Tim Oliver; Breuer, Christian; Garnitz, Johanna; Grimme, Christian; Hristov, Atanas; Hristov, Nikolay; Meister, Wolfgang; Reif, Magnus; Schröter, Felix; Steiner, Andreas; Wohlrabe, Klaus; Wolf, Anna

    2015-01-01

    Am 9. Dezember 2015 stellte das ifo Institut seine Prognose für die Jahre 2015, 2016 und 2017 vor. Der verhaltene Aufschwung, in dem sich die deutsche Wirtschaft seit einiger Zeit befindet, wird sich fortsetzen. In diesem Jahr dürfte das reale Bruttoinlandsprodukt um 1,7% zulegen, bevor es im kommenden Jahr vor­aussichtlich um 1,9% steigt. Im Jahr 2017 dürfte die Zuwachsrate dann wieder auf 1,7% zurückgehen. Der private Konsum wird weiterhin die Stütze des Aufschwungs bleiben, der durch die e...

  14. Helium cooling of fusion reactors

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Baxi, C.; Bourque, R.; Dahms, C.; Inamati, S.; Ryder, R.; Sager, G.; Schleicher, R.

    1994-01-01

    On the basis of worldwide design experience and in coordination with the evolution of the International Thermonuclear Experimental Reactor (ITER) program, the application of helium as a coolant for fusion appears to be at the verge of a transition from conceptual design to engineering development. This paper presents a review of the use of helium as the coolant for fusion reactor blanket and divertor designs. The concept of a high-pressure helium cooling radial plate design was studied for both ITER and PULSAR. These designs can resolve many engineering issues, and can help with reaching the goals of low activation and high performance designs. The combination of helium cooling, advanced low-activation materials, and gas turbine technology may permit high thermal efficiency and reduced costs, resulting in the environmental advantages and competitive economics required to make fusion a 21st century power source. ((orig.))

  15. Communication from ST Group

    CERN Multimedia

    TS Department

    2008-01-01

    In order to prepare the organization of the Open Days, please note that FM Group will not able to take into account either specific requests for waste collection from 2nd to 6th of April, either removal or PC transport requests between the 31st and the 11th of March. We kindly ask you to plan the collection of any type of waste and the urgent transport of office furniture or PC before the 31st of March. Waste collection requests shall be formulated contacting FM Support at 77777 or at the email address mailto:Fm.Support@cern.ch; removal of office furniture or PC transport requests must be made using the EDH ‘Transport request’ form selecting the "Removals" or the "PC transport" category from the drop-down menu. For any question concerning the waste sorting, please consult the following web address: http://dechets-waste.web.cern.ch/dechets-waste/. Thank you for your understanding and collaboration. TS/FM Group

  16. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    The method of operating a water-cooled neutronic reactor having a graphite moderator is described which comprises flowing a gaseous mixture of carbon dioxide and helium, in which the helium comprises 40--60 volume percent of the mixture, in contact with the graphite moderator. 2 claims, 4 figures

  17. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

  18. Reactor facility

    International Nuclear Information System (INIS)

    Suzuki, Hiroaki; Murase, Michio; Yokomizo, Osamu.

    1997-01-01

    The present invention provides a BWR type reactor facility capable of suppressing the amount of steams generated by the mutual effect of a failed reactor core and coolants upon occurrence of an imaginal accident, and not requiring spacial countermeasures for enhancing the pressure resistance of the container vessel. Namely, a means for supplying cooling water at a temperature not lower by 30degC than the saturated temperature corresponding to the inner pressure of the containing vessel upon occurrence of an accident is disposed to a lower dry well below the pressure vessel. As a result, upon occurrence of such an accident that the reactor core should be melted and flown downward of the pressure vessel, when cooling water at a temperature not lower than the saturated temperature, for example, cooling water at 100degC or higher is supplied to the lower dry well, abrupt generation of steams by the mutual effect of the failed reactor core and cooling water is scarcely caused compared with a case of supplying cooling water at a temperature lower than the saturation temperature by 30degC or more. Accordingly, the amount of steams to be generated can be suppressed, and special countermeasure is no more necessary for enhancing the pressure resistance of the container vessel is no more necessary. (I.S.)

  19. Nuclear reactor

    International Nuclear Information System (INIS)

    Gilroy, J.E.

    1980-01-01

    An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

  20. Reactor licensing

    International Nuclear Information System (INIS)

    Harvie, J.D.

    2002-01-01

    This presentation discusses reactor licensing and includes the legislative basis for licensing, other relevant legislation , the purpose of the Nuclear Safety and Control Act, important regulations, regulatory document, policies, and standards. It also discusses the role of the CNSC, its mandate and safety philosophy

  1. Nuclear reactor

    International Nuclear Information System (INIS)

    Hattori, Sadao; Sekine, Katsuhisa.

    1987-01-01

    Purpose: To decrease the thickness of a reactor container and reduce the height and the height and plate thickness of a roof slab without using mechanical vibration stoppers. Constitution: Earthquake proofness is improved by filling fluids such as liquid metal between a reactor container and a secondary container and connecting the outer surface of the reactor container with the inner surface of the secondary container by means of bellows. That is, for the horizontal seismic vibrations, horizontal loads can be supported by the secondary container without providing mechanical vibration stoppers to the reactor container and the wall thickness can be reduced thereby enabling to simplify thermal insulation structure for the reduction of thermal stresses. Further, for the vertical seismic vibrations, verical loads can be transmitted to the secondary container thereby enabling to reduce the wall thickness in the same manner as for the horizontal load. By the effect of transferring the point of action of the container load applied to the roof slab to the outer circumferential portion, the intended purpose can be attained and, in addition, the radiation dose rate at the upper surface of the roof slab can be decreased. (Kamimura, M.)

  2. Reactor system

    International Nuclear Information System (INIS)

    Miyano, Hiroshi; Narabayashi, Naoshi.

    1990-01-01

    The represent invention concerns a reactor system with improved water injection means to a pressure vessel of a BWR type reactor. A steam pump is connected to a heat removing system pipeline, a high pressure water injection system pipeline and a low pressure water injection system pipeline for injecting water into the pressure vessel. A pump actuation pipeline is disposed being branched from a main steam pump or a steam relieaf pipeline system, through which steams are supplied to actuate the steam pump and supply cooling water into the pressure vessel thereby cooling the reactor core. The steam pump converts the heat energy into the kinetic energy and elevates the pressure of water to a level higher than the pressure of the steams supplied by way of a pressure-elevating diffuser. Cooling water can be supplied to the pressure vessel by the pressure elevation. This can surely inject cooling water into the pressure vessel upon loss of coolant accident or in a case if reactor scram is necessary, without using an additional power source. (I.N.)

  3. Reactor core

    International Nuclear Information System (INIS)

    Matsuura, Tetsuaki; Nomura, Teiji; Tokunaga, Kensuke; Okuda, Shin-ichi

    1990-01-01

    Fuel assemblies in the portions where the gradient of fast neutron fluxes between two opposing faces of a channel box is great are kept loaded at the outermost peripheral position of the reactor core also in the second operation cycle in the order to prevent interference between a control rod and the channel box due to bending deformation of the channel box. Further, the fuel assemblies in the second row from the outer most periphery in the first operation cycle are also kept loaded at the second row in the second operation cycle. Since the gradient of the fast neutrons in the reactor core is especially great at the outer circumference of the reactor core, the channel box at the outer circumference is bent such that the surface facing to the center of the reactor core is convexed and the channel box in the second row is also bent to the identical direction, the insertion of the control rod is not interfered. Further, if the positions for the fuels at the outermost periphery and the fuels in the second row are not altered in the second operation cycle, the gaps are not reduced to prevent the interference between the control rod and the channel box. (N.H.)

  4. New about research reactors

    International Nuclear Information System (INIS)

    Egorenkov, P.M.

    2001-01-01

    The multi-purpose research reactor MAPLE (Canada) and concept of new reactor MAPLE-CNF as will substitute the known Canadian research reactor NRU are described. New reactor will be used as contributor for investigations into materials, neutron beams and further developments for the CANDU type reactor. The Budapest research reactor (BRR) and its application after the last reconstruction are considered also [ru

  5. Proceeding of 29th domestic symposium on computational science and nuclear energy in the 21st century

    International Nuclear Information System (INIS)

    2001-10-01

    As the 29th domestic symposium of Atomic Energy Research Committee, the Japan Welding Engineering Society, the symposium was held titled as Computational science and nuclear energy in the 21st century'. Keynote speech was delivered titled as 'Nuclear power plants safety secured by computational science in the 21st century'. Three speakers gave lectures titled as 'Materials design and computational science', 'Development of advanced reactor in the 21st century' and 'Application of computational science to operation and maintenance management of plants'. Lectures held panel discussion titled as 'Computational science and nuclear energy in the 21st century'. (T. Tanaka)

  6. Geology and hydrology of the Fort Belknap Indian Reservation, Montana

    Science.gov (United States)

    Alverson, Douglas C.

    1965-01-01

    The Fort Belknap Indian Reservation includes an area of 970 square miles in north-central Montana. At its north edge is the Milk River valley, which is underlain by Recent alluvium of the Milk River, glacial deposits, and alluvial deposits of the preglacial Missouri River, which carved and occupied this valley before the Pleistocene Epoch. Rising gently to the south is an undulating glaciated plain broken only by three small syenite porphyry intrusions. Underlying the glacial till of the plain are Upper Cretaceous shale and sandstone of the Bearpaw and Judith River Formations. At the south end of the reservation, 40 miles from the Milk River, an intrusion of syenite porphyry in Tertiary time uplifted, tilted, and exposed the succession of sedimentary rocks overlying the Precambrian metamorphic basement. The sedimentary rocks include 1,000 feet of sandstone and shale of Cambrian age; 2,000 feet of limestone and dolomite of Ordovician, Devonian, and Mississippian age; 400 feet of shale and limestone of Jurassic age; and 3,500 feet of sandstone, siltstone, and shale of Cretaceous age. Extensive gravel terraces of Tertiary and Quaternary age bevel the upturned bedrock formations exposed around the Little Rocky Mountains. Ground water under water-table conditions is obtained at present from alluvium, glaciofluvial deposits, and the Judith River Formation. The water table ranges in depth from a few feet beneath the surface in the Milk River valley alluvium to more than 100 feet deep in the Judith River Formation. Yields to wells are generally low but adequate for domestic and stock-watering use. Quality of the water ranges from highly mineralized and unusable to excellent; many wells in the Milk River valley have been abandoned because of the alkalinity of their water. Potential sources of additional ground-water supplies are the alluvial gravel of creeks issuing from the Little Rocky Mountains and some extensive areas of terrace gravel. The uplift and tilting of the

  7. Habitat restoration/enhancement Fort Hall Reservation : 2001 annual report

    International Nuclear Information System (INIS)

    Moser, David C.

    2003-01-01

    Habitat enhancement, protection and monitoring were the focus of the Resident Fisheries Program during 2001. Enhancement and protection included sloping, fencing and planting willows at sites on Diggie Creek, Clear Creek and Spring Creek. In addition, many previously constructed instream structures (rock barbs and wing dams) were repaired throughout the Fort Hall Indian Reservation (Reservation). In 2001, exclosure fences were erected on Diggie Creek (250 m barbed wire; (70 m jack), Wood Creek (500 m jack), Clear Creek (20 m jack), Ross Fork Creek (200 m jack), West Fork Creek (200 m jack)) and the Portneuf River (1 km barbed wire; 100 m jack). Jack and rail exclosure fences that had deteriorated over the past ten years were repaired at numerous areas throughout the Reservation. Physical sampling during 2001 included sediment and depth surveys (SADMS) in Big Jimmy Creek and Diggie Creek. SADMS, used to track changes in channel morphology and specifically track movements of silt through Bottoms stream systems were completed for eight and nine strata in the Big Jimmy and Diggie Creek, respectively. Baseline SADM data was collected in Diggie Creek to monitor the effects of bank sloping and revegetation on channel morphology and sediment levels through time. Water temperature was monitored (hourly) in Spring Creek, Clear Creek, Ross Fork Creek and Big Jimmy Creek. Biotic sampling included invertebrate sampling in the 200 and 300 series of Clear Creek. Fish population densities and biomass were sampled in Clear Creek 200 and 300 series. Sampling protocols were identical to methods used in past years. Numbers of fish in Clear Creek 300 series remained similar to 2000 while numbers of fish in Clear Creek 200 series dropped to near pre project levels. Salmonid fry densities were monitored near Broncho Bridge and were significantly higher than 2000. A mark-recapture study was initiated in spring 2001 to estimate numbers of spawning adults using the Head End of Spring Creek

  8. Reactivity estimation for subcritical and critical reactors

    International Nuclear Information System (INIS)

    Benhaim A; Bellino P; Gomez A

    2012-01-01

    We developed a digital reactimeter that works in both current and pulse mode. This reactimeter will allow to estimate the reactivity of the reactor at any state. We st obtained for the measurements taken in the experimental reactor RA-1 the reactivity around the critical state without a neutron source. Measurements were made using simultaneously a compensated ionization chamber and a 3He proportional counter. The results were compared with the ones obtained from the digital reactimeter of reference with matching results within the experimental errors (author)

  9. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

    1994-01-01

    In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

  10. ACHP | News | St. Elizabeths Programmatic Agreement Signed

    Science.gov (United States)

    Search skip specific nav links Home arrow News arrow St. Elizabeths Programmatic Agreement Signed St redevelopment of the St. Elizabeths West Campus, which is part of the St. Elizabeths National Historic Landmark this project, due to the historic significance of the NHL. GSA's client for the St. Elizabeths

  11. Surgical risk factors and maxillary nerve function after le fort I osteotomy

    DEFF Research Database (Denmark)

    Thygesen, Torben Henrik; Jensen, Allan Bardow; Norholt, SE

    2009-01-01

    PURPOSE: Data on intraoperative risk factors for long-term postoperative complications after Le Fort I osteotomy (LFO) are limited. The aim of this study was to describe prospectively the overall postoperative changes in maxillary nerve function after LFO, and to correlate these changes with a nu......PURPOSE: Data on intraoperative risk factors for long-term postoperative complications after Le Fort I osteotomy (LFO) are limited. The aim of this study was to describe prospectively the overall postoperative changes in maxillary nerve function after LFO, and to correlate these changes...

  12. Wind resource assessment and wind energy system cost analysis: Fort Huachuca, Arizona

    Energy Technology Data Exchange (ETDEWEB)

    Olsen, T.L. [Tim Olsen Consulting, Denver, CO (United States); McKenna, E. [National Renewable Energy Lab., Golden, CO (United States)

    1997-12-01

    The objective of this joint DOE and National Renewable Energy Laboratory (NREL) Strategic Environmental Research and Development Program (SERDP) project is to determine whether wind turbines can reduce costs by providing power to US military facilities in high wind areas. In support of this objective, one year of data on the wind resources at several Fort Huachuca sites was collected. The wind resource data were analyzed and used as input to an economic study for a wind energy installation at Fort Huachuca. The results of this wind energy feasibility study are presented in the report.

  13. Slaget ved Vejle og andre fortællinger fra Jyske Bank

    DEFF Research Database (Denmark)

    Albrechtsen, Charlotte

    Storytelling som ledelsesværktøj er en form for retorik idet formålet med at bruge fortællinger i kommunikationen fra ledelse til medarbejdere er at påvirke modtagerne/medarbejderne. Imidlertid er refleksioner over modtagerinstansen så godt som fraværende både i den populære debat om storytelling...... og i den eksisterende forskning i emnet. Foruden at introducere til forskningen i storytelling præsenterer artiklens forfatter, som er ph.d.-studerende, en modtagerorienteret analyse af en fortælling fra Jyske Bank....

  14. Nuclear reactor

    International Nuclear Information System (INIS)

    Gibbons, J.F.; McLaughlin, D.J.

    1978-01-01

    In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Sasaki, Tomozo.

    1987-01-01

    Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)

  16. Nuclear reactor

    International Nuclear Information System (INIS)

    Sakurai, Mikio; Yamauchi, Koki.

    1983-01-01

    Purpose: To improve the channel stability and the reactor core stability in a spontaneous circulation state of coolants. Constitution: A reactor core stabilizing device comprising a differential pressure automatic ON-OFF valve is disposed between each of a plurality of jet pumps arranged on a pump deck. The stabilizing device comprises a piston exerted with a pressure on the lower side of the pump deck by way of a pipeway and a valve for flowing coolants through the bypass opening disposed to the pump deck by the opening and closure of the valve ON-OFF. In a case where the jet pumps are stopped, since the differential pressure between the upper and the lower sides of the pump deck is removed, the valve lowers gravitationally into an opened state, whereby the coolants flow through the bypass opening to increase the spontaneous circulation amount thereby improve the stability. (Yoshino, Y.)

  17. Nuclear reactor

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.; Struensee, S.

    1976-01-01

    The invention concerns the use of burnable poisons in a nuclear reactor, especially in PWRs, in order to improve the controllability of the reactor. An unsymmetrical arrangement in the lattice is provided, if necessary also by insertion of special rods for these additions. It is proposed to arrange the burnable poisons in fuel elements taken over from a previous burn-up cycle and to distribute them, going out from the side facing the control rods, over not more than 20% of the lenth of the fuel elements. It seems sufficient, for the burnable poisons to bind an initial reactivity of only 0.1% and to become ineffective after normal operation of 3 to 4 months. (ORU) [de

  18. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  19. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  20. 77 FR 37318 - Eighth Coast Guard District Annual Safety Zones; Sound of Independence; Santa Rosa Sound; Fort...

    Science.gov (United States)

    2012-06-21

    ...-AA00 Eighth Coast Guard District Annual Safety Zones; Sound of Independence; Santa Rosa Sound; Fort... Coast Guard will enforce a Safety Zone for the Sound of Independence event in the Santa Rosa Sound, Fort... during the Sound of Independence. During the enforcement period, entry into, transiting or anchoring in...

  1. 78 FR 78380 - Notice of Inventory Completion: U.S. Department of the Interior, National Park Service, Fort...

    Science.gov (United States)

    2013-12-26

    ... completion of an inventory of human remains under the control of Fort Bowie National Historic Site, Bowie, AZ....R50000] Notice of Inventory Completion: U.S. Department of the Interior, National Park Service, Fort... completed an inventory of human remains, in consultation with the appropriate Indian tribes or Native...

  2. Fra erfaringer til betydninger: Tolkning af fortællinger om eksamensgruppebegivenheder fra folkeskolelærerstuderende ved Aalborg Seminarium

    DEFF Research Database (Denmark)

    Silleborg, Ellen

    2004-01-01

    karakter. Denne situation giveruoverensstemmelser i de studerendes følelsesliv, og hovedparten af fortællingerne afspejler en i mange henseender konfliktfyldt eksamensgruppeproces. I fortællingernes betydninger ses en loyal men privatiseret etik, hvor ansvarlighed bliver til selvskyld. Alle implicerede...

  3. 77 FR 57112 - Notice of Inventory Completion: U.S. Department of Defense, Army, Fort Sill Museum, Lawton, OK

    Science.gov (United States)

    2012-09-17

    ... Landmark and Museum, U.S. Army Fires Center of Excellence, Fort Sill, OK 73503, telephone (580) 442-6570... trapping, 3 metal rings, 2 metal rivets, 17 metal nails, 53 metal bracelets, 1 metal pail, 1,500 glass... A. Neel, Director, Fort Sill National Historic Landmark and Museum, U.S. Army Fires Center of...

  4. Airborne electromagnetic data and processing within Leach Lake Basin, Fort Irwin, California: Chapter G in Geology and geophysics applied to groundwater hydrology at Fort Irwin, California

    Science.gov (United States)

    Bedrosian, Paul A.; Ball, Lyndsay B.; Bloss, Benjamin R.; Buesch, David C.

    2014-01-01

    From December 2010 to January 2011, the U.S. Geological Survey conducted airborne electromagnetic and magnetic surveys of Leach Lake Basin within the National Training Center, Fort Irwin, California. These data were collected to characterize the subsurface and provide information needed to understand and manage groundwater resources within Fort Irwin. A resistivity stratigraphy was developed using ground-based time-domain electromagnetic soundings together with laboratory resistivity measurements on hand samples and borehole geophysical logs from nearby basins. This report releases data associated with the airborne surveys, as well as resistivity cross-sections and depth slices derived from inversion of the airborne electromagnetic data. The resulting resistivity models confirm and add to the geologic framework, constrain the hydrostratigraphy and the depth to basement, and reveal the distribution of faults and folds within the basin.

  5. Neutronic reactor

    International Nuclear Information System (INIS)

    Lewis, W.R.

    1978-01-01

    Disclosed is a graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels

  6. Nuclear reactors

    International Nuclear Information System (INIS)

    Humphreys, P.; Davidson, D.F.; Thatcher, G.

    1980-01-01

    The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

  7. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

    Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

  8. Reactor building

    International Nuclear Information System (INIS)

    Ebata, Sakae.

    1990-01-01

    At least one valve rack is disposed in a reactor building, on which pipeways to a main closure valve, valves and bypasses of turbines are placed and contained. The valve rack is fixed to the main body of the building or to a base mat. Since the reactor building is designed as class A earthquake-proofness and for maintaining the S 1 function, the valve rack can be fixed to the building main body or to the base mat. With such a constitution, the portions for maintaining the S 1 function are concentrated to the reactor building. As a result, the dispersion of structures of earthquake-proof portion corresponding to the reference earthquake vibration S 1 can be prevented. Accordingly, the conditions for the earthquake-proof design of the turbine building and the turbine/electric generator supporting rack are defined as only the class B earthquake-proof design conditions. In view of the above, the amount of building materials can be saved and the time for construction can be shortened. (I.S.)

  9. Nuclear reactors

    International Nuclear Information System (INIS)

    Yoshioka, Michiko.

    1985-01-01

    Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

  10. MLR reactor

    International Nuclear Information System (INIS)

    Ryazantsev, E.P.; Egorenkov, P.M.; Nasonov, V.A.; Smimov, A.M.; Taliev, A.V.; Gromov, B.F.; Kousin, V.V.; Lantsov, M.N.; Radchenko, V.P.; Sharapov, V.N.

    1998-01-01

    The Material Testing Loop Reactor (MLR) development was commenced in 1991 with the aim of updating and widening Russia's experimental base to validate the selected directions of further progress of the nuclear power industry in Russia and to enhance its reliability and safety. The MLR reactor is the pool-type one. As coolant it applies light water and as side reflector beryllium. The direction of water circulation in the core is upward. The core comprises 30 FA arranged as hexagonal lattice with the 90-95 mm pitch. The central materials channel and six loop channels are sited in the core. The reflector includes up to 11 loop channels. The reactor power is 100 MW. The average power density of the core is 0.4 MW/I (maximal value 1.0 MW/l). The maximum neutron flux density is 7.10 14 n/cm 2 s in the core (E>0.1 MeV), and 5.10 14 n/cm 2 s in the reflector (E<0.625 eV). In 1995 due to the lack of funding the MLR designing was suspended. (author)

  11. 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.)

  12. Reactor container

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

    1990-01-01

    In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

  13. Reactor core in FBR type reactor

    International Nuclear Information System (INIS)

    Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

    1989-01-01

    In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

  14. Calculation of induced activity in the V-230 reactor

    International Nuclear Information System (INIS)

    Bouhahhane, A.; Farkas, G.

    2013-01-01

    In this paper, we focused on the calculation of the neutron induced activity of nuclear reactor components for decommissioning purposes. The results confirm, that the most important radionuclides in the reactor components dismantling process are 55 Fe (1 st decade), 60 Co (10 - 50 y) and 63 Ni (during the whole process). Another aim of this paper was to refer to the possibility to improve the accuracy of the calculations using continuous energy Monte Carlo methods. (authors)

  15. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  16. Comparisons among different development ways of advanced reactors in China

    International Nuclear Information System (INIS)

    Guo Xingqu; Lin Jianwen; Wang Ruoli

    1992-03-01

    For the development of nuclear energy in the 21st century, China will select a new type reactor to develop, which will have higher fuel efficiency, high safety and better economics. The selection is among the types of FBR (fast breeder reactor), HTGR (high temperature gas-cooled reactor) and FFHR (fusion-fission hybrid reactor). Since the evaluation of advanced reactors involves many uncertain factors and the difficulty of quantization, both the AHP (analytic hierarchy process) method and expert consultation are adopted. Four aspects are taken in the norm system of AHP, i.e. safety, maturity of technology, economy and appropriateness. By using questionnaire method to experts and studying related documents, five types of advanced reactor are selected, i.e. oxide fueled FBR, metal fueled FBR, uranium fueled HTGR, U-Th fueled HTGR and FFBR. Their evaluation parameters are a comprehensively assessed and sorted. About 130 experts and professors who have been working in the research institutes and government agencies of nuclear field are asked to give their comments on the development of advanced reactors. The response rate of questionnaires is 86%, and the data collected are processed by computers. From the evaluation result of AHP method and expert consultation of the fast breeder reactor, especially, the metal fueled FBR, should have the priority in nuclear energy development in the 21st century in China

  17. Molten salt reactors: reactor cores

    International Nuclear Information System (INIS)

    1983-01-01

    In this critical analysis of the MSBR I project are examined the problems concerning the reactor core. Advantages of breeding depend essentially upon solutions to technological problems like continuous reprocessing or graphite behavior under neutron irradiation. Graphite deformation, moderator unloading, control rods and core instrumentation require more studies. Neutronics of the core, influence of core geometry and salt composition, fuel evolution, and thermohydraulics are reviewed [fr

  18. Increased SRP reactor power

    International Nuclear Information System (INIS)

    MacAfee, I.M.

    1983-01-01

    Major changes in the current reactor hydraulic systems could be made to achieve a total of about 1500 MW increase of reactor power for P, K, and C reactors. The changes would be to install new, larger heat exchangers in the reactor buildings to increase heat transfer area about 24%, to increase H 2 O flow about 30% per reactor, to increase D 2 O flow 15 to 18% per reactor, and increase reactor blanket gas pressure from 5 psig to 10 psig. The increased reactor power is possible because of reduced inlet temperature of reactor coolant, increased heat removal capacity, and increased operating pressure (larger margin from boiling). The 23% reactor power increase, after adjustment for increased off-line time for reactor reloading, will provide a 15% increase of production from P, K, and C reactors. Restart of L Reactor would increase SRP production 33%

  19. Nuclear research reactors

    International Nuclear Information System (INIS)

    1985-01-01

    It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

  20. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    Baeten, P.

    2006-01-01

    The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

  1. 78 FR 53494 - Dam Safety Modifications at Cherokee, Fort Loudoun, Tellico, and Watts Bar Dams

    Science.gov (United States)

    2013-08-29

    ... Bar Dams AGENCY: Tennessee Valley Authority. ACTION: Issuance of Record of Decision. SUMMARY: This... the dam safety modifications at Cherokee, Fort Loudoun, Tellico, and Watts Bar Dams. The notice of... Loudoun, Tellico, and Watts Bar Dams was published in the Federal Register on May 31, 2013. This...

  2. Craniofacial stability in patients with Crouzon or Apert syndrome after Le Fort III distraction osteogenesis

    NARCIS (Netherlands)

    Reitsma, J.H.; Ongkosuwito, E.M.; Buschang, P.H.; van Adrichem, L.N.A.; Prahl-Andersen, B.

    2013-01-01

    Objective: Le Fort III osteotomy with distraction osteogenesis (DO) is used to improve the retruded midface in patients with Crouzon or Apert syndrome. This study aimed to evaluate sagittal and vertical preoperative and postoperative cephalometric changes of DO of the midface in patients with

  3. Craniofacial stability in patients with crouzon or apert syndrome after le fort III distraction osteogenesis

    NARCIS (Netherlands)

    J.H. Reitsma (Jacobus Harmen); E.M. Ongkosuwito (Edwin); P.H. Buschang (Peter); L.N.A. V Adrichem (Léon); B. Prahl-Andersen (Birte)

    2013-01-01

    textabstractObjective: Le Fort III osteotomy with distraction osteogenesis (DO) is used to improve the retruded midface in patients with Crouzon or Apert syndrome. This study aimed to evaluate sagittal and vertical preoperative and postoperative cephalometric changes of DO of the midface in patients

  4. Management of Groin Abcess with Flaminal Forte and KerraMax Care

    Directory of Open Access Journals (Sweden)

    Maggie Pugh

    2016-04-01

    Full Text Available The patient’s dressing plan using Flaminal Forte and KerraMax Care successfully managed the complexities of his wound, absorbing exudate, reducing pain on dressing, malodour and wound bioburden. Moreover, the plan encouraged patient concordance, reduced nursing consultation time and subsequently altered treatment plans for our patients with abscesses

  5. 76 FR 77684 - Establishment of the Fort Ross-Seaview Viticultural Area

    Science.gov (United States)

    2011-12-14

    ...; Treasury decision. SUMMARY: This Treasury decision establishes the 27,500-acre ``Fort Ross-Seaview... may purchase. DATES: Effective Date: January 13, 2012. FOR FURTHER INFORMATION CONTACT: Elisabeth C... may purchase. Establishment of a viticultural area is neither an approval nor an endorsement by TTB of...

  6. Notes on Chameleons IV. A New Chameleon, from the Miocene of Fort Ternan, Kenya (Chamaeleonidae, Reptilia)

    NARCIS (Netherlands)

    Hillenius, D.

    1978-01-01

    A new, fossil chameleon is described, † Chamaeleo intermedius, found on the surface at Fort Ternan, Kenya. Most probably it eroded from a layer of fossilized lahar, close to 14 million years old. † Chamaeleo intermedius possesses characters which still occur in recent chameleons, in fact it combines

  7. Wooden combs from the Roman fort at Vechten: the bodily appearance of soldiers

    NARCIS (Netherlands)

    Derks, A.M.J.; Vos, W.K.

    2010-01-01

    Abstract Excavations in the late 19th century and surveys carried out in the 1970s have produced 12 boxwood combs from the Roman fort at Vechten (NL). They are to be considered waste material that was dumped in the river Rhine which in the Roman period ran just north of the camp. In this article,

  8. Biology and ecology of sickleweed (Falcaria vulgaris) in the Fort Pierre National Grassland of South Dakota

    Science.gov (United States)

    Brian L. Korman

    2011-01-01

    In the last two decades the exotic plant sickleweed (Falcaria vulgaris Bernh., Apiaceae) has invaded, and come to dominate, large areas of the Fort Pierre National Grassland (FPNG) in central South Dakota, USA. Currently sickleweed is estimated to infest over 3200 ha of FPNG. The purpose of this study was to examine several of the biological and ecological traits that...

  9. Fort Hood Building and Landscape Inventory with WWII and Cold War Context

    Science.gov (United States)

    2007-03-01

    barracks, 1970s (NARA)........................................................... 112 Figure 37. Palmer Movie Theater (NARA...revised 1953) showing layout of Hood Village and trailer park (Fort Hood...arms ammunition storage building #92012 (ERDC-CERL, 2004). ......... 260 Figure 163: Radio reception building #92063 (ERDC-CERL, 2004

  10. 78 FR 4356 - Proposed Modification of the Dallas/Fort Worth Class B Airspace Area; TX

    Science.gov (United States)

    2013-01-22

    ... Dallas/Fort Worth International Airport (DFW) and Dallas Love Field Airport (DAL) within Class B airspace... acknowledge receipt of their comments on this action must submit with those comments a self-addressed, stamped... configuration has not kept pace with airport expansions and increasing operations and the current design makes...

  11. 77 FR 21448 - Security Zone; 2012 Fleet Week, Port Everglades, Fort Lauderdale, FL

    Science.gov (United States)

    2012-04-10

    ... Environmental Health Risks and Safety Risks. This rule is not an economically significant rule and does not create an environmental risk to health or risk to safety that may disproportionately affect children...-AA87 Security Zone; 2012 Fleet Week, Port Everglades, Fort Lauderdale, FL AGENCY: Coast Guard, DHS...

  12. Revisiting Pearson's climate and forest type studies on the Fort Valley Experimental Forest (P-53)

    Science.gov (United States)

    Joseph E. Crouse; Margaret M. Moore; Peter Z. Fule

    2008-01-01

    Five weather station sites were established in 1916 by Fort Valley personnel along an elevational gradient from the Experimental Station to near the top of the San Francisco Peaks to investigate the factors that controlled and limited forest types. The stations were located in the ponderosa pine, Douglas-fir, limber pine, Engelmann spruce, and Engelmann spruce/...

  13. The Fort Logan Lodge: Intentional Community for Chronic Mental Patients. Final Report.

    Science.gov (United States)

    Fort Logan Mental Health Center, Denver, CO.

    This report attempts to identify important variables affecting the success of the Lodge Program, affiliated with the Fort Logan Mental Health Center. The Lodge Program is a community based, group oriented, social and work program for the rehabilitation of the refractory, long stay mental patient. Findings reported include the following: (1) the…

  14. Historical Analysis of Land Cover/Condition Trends at Fort Bliss, Texas, Using Remotely Sensed Imagery

    National Research Council Canada - National Science Library

    Tweddale, Scott

    2001-01-01

    .... They need a cost-effective method of assessing and monitoring land condition. The objective of this research was to characterize the small scale, gross level change in land condition on a selected area of Fort Bliss over a 23-year period...

  15. 78 FR 3479 - Notice of Public Meeting of Fort Scott Council

    Science.gov (United States)

    2013-01-16

    ... submitted on cards that will be provided at the meeting, via mail to Laurie Fox, Presidio Trust, 103... stated prominently at the beginning of the comments. The Trust will make available for public inspection... PRESIDIO TRUST Notice of Public Meeting of Fort Scott Council AGENCY: The Presidio Trust. ACTION...

  16. 76 FR 71611 - Notice of Establishment of the Fort Winfield Scott Advisory Committee

    Science.gov (United States)

    2011-11-18

    ... (``Committee''). The Committee will advise the Executive Director of the Presidio Trust on matters pertaining... of once every three months. Nominations: The Presidio Trust will consider nominations of all... PRESIDIO TRUST Notice of Establishment of the Fort Winfield Scott Advisory Committee AGENCY: The...

  17. 78 FR 60929 - Notice of Public Meeting of the Fort Scott Council

    Science.gov (United States)

    2013-10-02

    .... Such requests must be stated prominently at the beginning of the comments. The Trust will make... PRESIDIO TRUST Notice of Public Meeting of the Fort Scott Council AGENCY: The Presidio Trust... Scott Council (Council) will be held from 10 a.m. to 12:30 p.m. on Thursday, October 17, 2013. The...

  18. Inicial nasal mucosa detachment using piezoelectric device in the Le Fort I osteotomy: A technical note

    OpenAIRE

    Shinohara, Elio Hitoshi; Kaba, Shajadi Carlos Pardo; Ruiz, Marcelo Martinson; Horikawa, Fernando Kendi

    2013-01-01

    In maxillary Le Fort I type osteotomy the detachment of the nasal mucosa should be done carefully. Piezoelectric surgery contributed much to increase the safety of osteotomies, despite the initial advantage of minimizing the risk of injury in nervous tissue, mainly in bilateral sagittal split osteotomy; we use the piezoelectric device for the initial detachment of the nasal mucosa in the maxillary osteotomy.

  19. 75 FR 52733 - Record of Decision (ROD) for Fort Bliss Army Growth and Force Structure Realignment

    Science.gov (United States)

    2010-08-27

    ... decrease in certain quality of life indicators (e.g., traffic, air quality). The ROD, FEIS and other environmental documents are available on the Fort Bliss Web site ( http://www.bliss.army.mil ) or at the... (Stationing Action Alternative 4); land use changes that allow fixed site bivouac areas, mission support...

  20. 75 FR 40034 - Northeastern Tributary Reservoirs Land Management Plan, Beaver Creek, Clear Creek, Boone, Fort...

    Science.gov (United States)

    2010-07-13

    ... TENNESSEE VALLEY AUTHORITY Northeastern Tributary Reservoirs Land Management Plan, Beaver Creek...-managed public land on Beaver Creek, Clear Creek, Boone, Fort Patrick Henry, South Holston, Watauga, and... Proposed Land Use Alternative) identified in the final environmental impact statement (FEIS). Under the...

  1. 77 FR 51064 - Huntington Foam LLC, Fort Smith, AR; Notice of Affirmative Determination Regarding Application...

    Science.gov (United States)

    2012-08-23

    ... DEPARTMENT OF LABOR Employment and Training Administration [TA-W-81,475] Huntington Foam LLC, Fort Smith, AR; Notice of Affirmative Determination Regarding Application for Reconsideration By application dated May 21, 2012, the State Workforce Office requested administrative reconsideration of the negative...

  2. Possible Location of Gaspar Dias Fort in Relation to the Present River Bank

    Digital Repository Service at National Institute of Oceanography (India)

    Mascarenhas, A.; Tripati, S.; ManiMurali, R.

    , it would be worthwhile to delineate the past river bank with respect to the present one, and to check whether any morphological changes occurred since then. Although the fort is marked by the river side in historical maps, the exact position of the shore...

  3. Evaluation of Eurasian Watermilfoil Control Techniques Using Aquatic Herbicides in Fort Peck Lake, Montana

    Science.gov (United States)

    2015-07-01

    Dredge Cut #2) are located immediately below Fort Peck Dam (Figure 4). The Dredge Cuts were formed by the excavation of soil for construction of the... Enviro -USA) consisted of 50, 6 m × 4.1 m deep sections. When sections were connected, a total length of 305 m was achieved. The top of the curtain was

  4. Assessment of loblolly pine decline and site conditions on Fort Benning Military Reservation, GA

    Science.gov (United States)

    Roger D. Menard; Lori G. Eckhardt; Nolan J. Hess

    2010-01-01

    A decline of loblolly pine (Pinus taeda L.), characterized by expanding areas of declining and dead trees, has become prevalent at Fort Benning, GA. A 3-year study was conducted to determine the kinds of fungi, insects, and site disturbances associated with this problem. The insects Dendroctonus terebrans, Hylastes salebrosus, H. tenuis, Pachylobius picivorus...

  5. Demonstration of Thermally Sprayed Metal and Polymer Coatings for Steel Structures at Fort Bragg, NC

    Science.gov (United States)

    2017-09-01

    ER D C/ CE RL T R- 17 -3 0 DoD Corrosion Prevention and Control Program Demonstration of Thermally Sprayed Metal and Polymer Coatings...and Polymer Coatings for Steel Structures at Fort Bragg, NC Final Report on Project F07-AR10 Larry D. Stephenson, Alfred D. Beitelman, Richard G...5 2.1.2 Thermoplastic polymer coating (flame spray

  6. Military vehicle trafficking impacts vegetation and soil bulk density at Fort Benning, Georgia

    Science.gov (United States)

    Potential increases in wind erosion that might be brought about by military vehicles travelling off-road during training are of concern to the United States military. Field studies were conducted in the summer of 2012 at Fort Benning, Georgia. The objective of the experiment was to assess the traffi...

  7. 75 FR 36371 - Draft Environmental Impact Statement Addressing Campus Development at Fort Meade, MD

    Science.gov (United States)

    2010-06-25

    ...'s (NSA) continually evolving requirements and for Intelligence Community use. The purpose of the..., or e-mail [email protected]nsa.gov . SUPPLEMENTARY INFORMATION: Background: The NSA is a tenant DOD agency on Fort Meade. NSA is a high-technology organization that is on the frontier of communications and data...

  8. 76 FR 22338 - Proposed Fort Ross-Seaview Viticultural Area; Comment Period Reopening

    Science.gov (United States)

    2011-04-21

    ... May 9, 2005, from all interested persons. In response to a request from a local wine industry member... the Fort Ross-Seaview viticultural area. Two local wine industry members supported the petition without qualification; a third industry member supported the viticultural area's establishment while...

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    Jolly, R.

    1979-01-01

    The support grid for the fuel rods of a liquid metal cooled fast breeder reactor has a regular hexagonal contour and contains a large number of unit cells arranged honeycomb fashion. The totality of these cells make up a hexagonal shape. The grid contains a number of strips of material, and there is a window in each of three sidewalls staggered by one sidewall. The other sidewalls have embossed protrusions, thus generating a guide lining or guide bead. The windows reduce the rigidity of the areas in the middle between the ends of the cells. (DG) [de

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    Anthony, A.J.; Gruber, E.A.

    1979-01-01

    A nuclear reactor with control rods in channels between fuel assemblies wherein the fuel assemblies incorporate guide rods which protrude outwardly into the control rod channels to prevent the control rods from engaging the fuel elements. The guide rods also extend back into the fuel assembly such that they are relatively rigid members. The guide rods are tied to the fuel assembly end or support plates and serve as structural members which are supported independently of the fuel element. Fuel element spacing and support means may be attached to the guide rods. 9 claims

  11. Maxillary distraction osteogenesis at Le Fort-I level induces bone apposition at infraorbital rim.

    Science.gov (United States)

    Rattan, Vidya; Jena, Ashok Kumar; Singh, Satinder Pal; Utreja, Ashok Kumar

    2014-09-01

    The aim of this study is to evaluate whether there is any remodeling of bone at infraorbital rim following maxillary distraction osteogenesis (DO) at Le Fort-I level. Twelve adult subjects in the age range of 17-21 years with complete unilateral cleft lip and palate underwent advancement of the maxilla by DO. The effect of maxillary DO on the infraorbital rim remodeling was evaluated from lateral cephalograms recorded prior to the DO (T0), at the end of DO (T1), and at least 2-years after the DO (T2) by Walker's analysis. The ANOVA and two-tailed t test were used and probability value (P value) 0.05 was considered as statistically significant level. There was anterior movement of maxilla by 9.22 ± 3.27 mm and 7.67 ± 3.99 mm at the end of immediate (T1) and long-term (T2) follow-up of maxillary DO, respectively. The Walker's analysis showed 1.49 ± 1.22 mm and 2.31 ± 1.81 mm anterior movement of the infraorbital margin (Orbitale point) at the end of T1 and T2, respectively (P distraction osteogenesis at Le Fort-I level induced significant bone apposition at infraorbital rim. Patients with mild midface hypoplasia who would otherwise may be candidates for osteotomy at Le Fort-II or Le Fort-III level may benefit from maxillary distraction at Le Fort-I level.

  12. Third party testing : new pilot facility for mining processes opens in Fort McKay

    International Nuclear Information System (INIS)

    Jaremko, D.

    2007-01-01

    Fort McKay lies 65 kilometres north of Fort McMurray, Alberta and is the centre of operational oilsands mining activity. As such, it was chosen for a pilot testing facility created by the Geneva-based SGS Group. The reputable facility provides an opportunity for mining producers to advance their processes, including environmental performance, by allowing them to test different processes on their own oilsands. The Northern Lights partnership, led by Synenco Energy, was the first client at the facility. Due to outsourcing, clients are not obligated to make substantial capital investment into in-house research. The Northern Lights partnership will be using the facility to test extraction processes on bitumen from its leases. Although the Fort McKay facility is SGS's first venture into the oilsands industry, it operates in more than 140 companies globally, including the mineral industry, and specializes in inspection, verification, testing and certification. SGS took the experience from its minerals extraction business to identify what could be done to help the oilsands industry by using best practices developed from global operations. The facility lies on the Fort McKay industrial park owned by the Fort McKay First Nation. An existing testing facility called McMurray Resources Research and Testing was expanded by the SGS Group to include environmental analysis capabilities. The modular units that lie on 6 acres include refrigerated ore storage to maintain ore integrity; modular ore and materials handling systems; extraction equipment; and, zero discharge process water and waste disposal systems. Froth treatment will be added in the near future to cover the entire upstream side of the mining processing business. A micro-upgrader might be added in the future to manufacture synthetic crude. 3 figs

  13. Third party testing : new pilot facility for mining processes opens in Fort McKay

    Energy Technology Data Exchange (ETDEWEB)

    Jaremko, D.

    2007-12-15

    Fort McKay lies 65 kilometres north of Fort McMurray, Alberta and is the centre of operational oilsands mining activity. As such, it was chosen for a pilot testing facility created by the Geneva-based SGS Group. The reputable facility provides an opportunity for mining producers to advance their processes, including environmental performance, by allowing them to test different processes on their own oilsands. The Northern Lights partnership, led by Synenco Energy, was the first client at the facility. Due to outsourcing, clients are not obligated to make substantial capital investment into in-house research. The Northern Lights partnership will be using the facility to test extraction processes on bitumen from its leases. Although the Fort McKay facility is SGS's first venture into the oilsands industry, it operates in more than 140 companies globally, including the mineral industry, and specializes in inspection, verification, testing and certification. SGS took the experience from its minerals extraction business to identify what could be done to help the oilsands industry by using best practices developed from global operations. The facility lies on the Fort McKay industrial park owned by the Fort McKay First Nation. An existing testing facility called McMurray Resources Research and Testing was expanded by the SGS Group to include environmental analysis capabilities. The modular units that lie on 6 acres include refrigerated ore storage to maintain ore integrity; modular ore and materials handling systems; extraction equipment; and, zero discharge process water and waste disposal systems. Froth treatment will be added in the near future to cover the entire upstream side of the mining processing business. A micro-upgrader might be added in the future to manufacture synthetic crude. 3 figs.

  14. Nuclear reactors

    International Nuclear Information System (INIS)

    Prescott, R.F.; George, B.V.; Baglin, C.J.

    1979-01-01

    In a nuclear reactor (e.g. one having coolant down-flow through a core to a hearth below) thermal insulation (e.g. of a floor of the hearth) comprises a layer of bricks and a layer of tiles thereon, with smaller clearances between the tiles than between the bricks but with the bricks being of reduced cross-section immediately adjacent the tiles so as to be surrounded by interconnected passages, of relatively large dimensions, constituting a continuous chamber extending behind the layer of tiles. By this arrangement, lateral coolant flow in the inter-brick clearances is much reduced. The reactor core is preferably formed of hexagonal columns, supported on diamond-shaped plates each supported on a pillar resting on one of the hearth-floor tiles. Each plate has an internal duct, four upper channels connecting the duct with coolant ducts in four core columns supported by the plate, and lower channels connecting the duct to a downwardly-open recess common to three plates, grouped to form a hexagon, at their mutually-adjacent corners. This provides mixing, and temperature-averaging, of coolant from twelve columns

  15. 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.)

  16. Reactor monitor

    International Nuclear Information System (INIS)

    Takada, Tamotsu.

    1992-01-01

    The device of the present invention monitors a reactor so that each of the operations for the relocation of fuel assemblies and the withdrawal and the insertion of control rods upon exchange of fuel assemblies and control rods in the reactor. That is, when an operator conducts relocating operation by way of a fuel assembly operation section, the device of the present invention judges whether the operation indication is adequate or not, based on the information of control rod arrangement in a control rod memory section. When the operation indication is wrong, a stop signal is sent to a fuel assembly relocating device. Further, when the operator conducts control rod operation by way of a control rod operation section, the device of the present invention judges in the control rod withdrawal judging section, as to whether the operation indication given by the operator is adequate or not by comparing it with fuel assembly arrangement information. When the operation indication is wrong, a stop signal is sent to control rod drives. With such procedures, increase of nuclear heating upon occurrence of erroneous operation can be prevented. (I.S.)

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Matheson, J.E.

    1983-01-01

    A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

  18. Reactor core fuel management

    International Nuclear Information System (INIS)

    Silvennoinen, P.

    1976-01-01

    The subject is covered in chapters, entitled: concepts of reactor physics; neutron diffusion; core heat transfer; reactivity; reactor operation; variables of core management; computer code modules; alternative reactor concepts; methods of optimization; general system aspects. (U.K.)

  19. Hybrid adsorptive membrane reactor

    Science.gov (United States)

    Tsotsis, Theodore T [Huntington Beach, CA; Sahimi, Muhammad [Altadena, CA; Fayyaz-Najafi, Babak [Richmond, CA; Harale, Aadesh [Los Angeles, CA; Park, Byoung-Gi [Yeosu, KR; Liu, Paul K. T. [Lafayette Hill, PA

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  20. Reactor outage schedule (tentative)

    Energy Technology Data Exchange (ETDEWEB)

    Walton, R.P.

    1969-11-01

    This single page document is the November 1, 1969 reactor refueling outage schedule for the Hanford Production Reactor. It also contains data on the amounts and types of fuels to be loaded and relocated in the production reactor.