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

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

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

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

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

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

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

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.)

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.)

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.

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

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

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

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

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

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.

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.)

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)

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

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

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)

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

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

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.

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

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

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

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

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

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)

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)

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)

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

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)

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

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)

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)

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.

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

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

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.)

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.

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

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

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

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

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.

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.

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

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

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

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)

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

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.

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

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

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

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

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

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

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.

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.

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

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

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

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

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.

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

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

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

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

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

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

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

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

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.

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

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.)

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

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

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

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

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.)

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)

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.)

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

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

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

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)

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

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.

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

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)

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

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)

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

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.

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)

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

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

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.

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

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

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

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

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.

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

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

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

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...

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.

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.

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...

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

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

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....

Remote inspection of the IFSF spent fuel storage rack

International Nuclear Information System (INIS)

Uldrich, E.D.

1996-01-01

The Irradiated Fuel Storage Facility (IFSF) is a dry storage facility for spent nuclear fuels located at the Idaho Chemical Processing Plant; it was constructed in the 1970's specifically for the Fort Saint Vrain spent reactor fuels. Currently, it is being used for various spent fuels. It was not known if IFSF would met current DOE seismic criteria, so re-analysis was started, with the rack being analyzed first. The rack was inspected to determine the as-built condition. LazrLyne and VideoRuler were used in lieu of using a tape measure with the camera. It was concluded that when a visual inspection shows widely varying weld sizes, the engineer has to use all resources available to determine the most probable specified weld sizes

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.

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)

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

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

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

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

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.

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.

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)

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

Humin to Human: Organic carbon, sediment, and water fluxes along river corridors in a changing world

Energy Technology Data Exchange (ETDEWEB)

Sutfin, Nicholas Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-11-20

This is a presentation with slides on What does it mean to be human? ...humin?; River flow and Hydrographs; Snake River altered hydrograph (Marston et al., 2005); Carbon dynamics are important in rivers; Rivers and streams as carbon sink; Reservoirs for organic carbon; Study sites in Colorado; River morphology; Soil sample collection; Surveys at RMNP; Soil organic carbon content at RMNP; Abandoned channels and Cutoffs; East River channel migration and erosion; Linking hydrology to floodplain sediment flux; Impact of Extreme Floods on Floodplain Sediment; Channel Geometry: RMNP; Beavers dams and multithread channels; Geomorphology and carbon in N. St. Vrain Creek; Geomorphology and carbon along the East River; Geomorphology and carbon in N. St. Vrain Creek; San Marcos River, etc.

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)

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...

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

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

STAT, GAPS, STRAIN, DRWDIM: a system of computer codes for analyzing HTGR fuel test element metrology data. User's manual

Energy Technology Data Exchange (ETDEWEB)

Saurwein, J.J.

1977-08-01

A system of computer codes has been developed to statistically reduce Peach Bottom fuel test element metrology data and to compare the material strains and fuel rod-fuel hole gaps computed from these data with HTGR design code predictions. The codes included in this system are STAT, STRAIN, GAPS, and DRWDIM. STAT statistically evaluates test element metrology data yielding fuel rod, fuel body, and sleeve irradiation-induced strains; fuel rod anisotropy; and additional data characterizing each analyzed fuel element. STRAIN compares test element fuel rod and fuel body irradiation-induced strains computed from metrology data with the corresponding design code predictions. GAPS compares test element fuel rod, fuel hole heat transfer gaps computed from metrology data with the corresponding design code predictions. DRWDIM plots the measured and predicted gaps and strains. Although specifically developed to expedite the analysis of Peach Bottom fuel test elements, this system can be applied, without extensive modification, to the analysis of Fort St. Vrain or other HTGR-type fuel test elements.

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...

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...

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

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

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

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)

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...

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.

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

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

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

Report to Congress on abnormal occurrences, July-September 1986

International Nuclear Information System (INIS)

1987-04-01

Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission determines to be significant from the standpoint of public health or safety and requires a quarterly report of such events to be made to Congress. This report covers the period from July 1 to September 30, 1986. The report states that for this reporting period, there were four abnormal occurrences at the nuclear power plants licensed to operate. The events were (1) a differential pressure switch problem in safety systems at LaSalle facility, (2) abnormal cooldown and depressurization transient at Catawba Unit 2, (3) significant safeguards deficiencies at Wolf Creek and Fort St. Vrain, and (4) significant deficiencies in access controls at River Bend Station. There was one abnormal occurrence at the other NRC licensees; it involved a therapeutic medical misadministration. There was one abnormal occurrence reported by an Agreement State; it involved a therapeutic medical misadministration. The report also contains information updating some previously reported abnormal occurrences

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.

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.

Lagrangian sampling for emerging contaminants through an urban stream corridor in Colorado

Science.gov (United States)

Brown, J.B.; Battaglin, W.A.; Zuellig, R.E.

2009-01-01

Recent national concerns regarding the environmental occurrence of emerging contaminants (ECs) have catalyzed a series of recent studies. Many ECs are released into the environment through discharges from wastewater treatment plants (WWTPs) and other sources. In 2005, the U.S. Geological Survey and the City of Longmont initiated an investigation of selected ECs in a 13.8-km reach of St. Vrain Creek, Colorado. Seven sites were sampled for ECs following a Lagrangian design; sites were located upstream, downstream, and in the outfall of the Longmont WWTP, and at the mouths of two tributaries, Left Hand Creek and Boulder Creek (which is influenced by multiple WWTP outfalls). Samples for 61 ECs in 16 chemical use categories were analyzed and 36 were detected in one or more samples. Of these, 16 have known or suspected endocrine-disrupting potential. At and downstream from the WWTP outfall, detergent metabolites, fire retardants, and steroids were detected at the highest concentrations, which commonly exceeded 1 ??g/l in 2005 and 2 ??g/l in 2006. Most individual ECs were measured at concentrations less than 2 ??g/l. The results indicate that outfalls from WWTPs are the largest but may not be the sole source of ECs in St. Vrain Creek. In 2005, high discharge was associated with fewer EC detections, lower total EC concentrations, and smaller EC loads in St. Vrain Creek and its tributaries as compared with 2006. EC behavior differed by individual compound, and some differences between sites could be attributed to analytical variability or to other factors such as physical or chemical characteristics or distance from contributing sources. Loads of some ECs, such as diethoxynonylphenol, accumulated or attenuated depending on location, discharge, and distance downstream from the WWTP, whereas others, such as bisphenol A, were largely conservative. The extent to which ECs in St. Vrain Creek affect native fish species and macroinvertebrate communities is unknown, but recent

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

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.

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.

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

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.

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.

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.)

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

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.

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

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.

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

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)

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

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.

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

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)

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

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...

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

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.

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

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...

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

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.

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

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…

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

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...

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...

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

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

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.

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.

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...

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...

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.

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...

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

Certification of the instructional competence of nuclear training specialists

International Nuclear Information System (INIS)

Wollert, T.N.

1990-01-01

This study was designed to identify the qualification requirements and the means to assess the unique knowledge and skills necessary to perform the instructional activities needed by nuclear training specialist at Fort Saint Vrain Nuclear Generating Station. A survey questionnaire with 233 task statements categorized into eleven duty areas was distributed to twenty-three nuclear training specialists at Fort Saint Vrain Nuclear Generating Station. On the basis of the data accumulated for this study, the researcher identified the following findings. A list of 158 task statements were identified as being relevant; this list was considered a core knowledge, skills, and abilities needed as a nuclear training specialist. The list consisted of ten duty areas which were relevant to the effective performance of a nuclear training specialist. Thirty-three task statements were identified as being relevant for the duty area Conductive Training. These were considered the core of knowledge, skills, and abilities needed in the development of the initial test instrument and the instructor classroom skills observation checklist. The significant correlation between the results of these two instruments, using a rank-order correlation coefficient, was interpreted by the researcher as indicating that the initial test instrument possessed concurrent validity. The researcher interpreted the reliability value as a positive indicator that the initial test instrument demonstrated internal consistency. It was concluded that it could be determined whether personnel possessed the level of competence needed to perform the instructional duties of a nuclear training specialist by using a written test. Data from this research supported the use of the initial test developed for this study as a valid means to certify nuclear training specialists for the duty area Conducting Training

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.

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.

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.

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

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

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)

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....

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

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.

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

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.

Revisão e análise cladística de Serdia Stål (Heteroptera, Pentatomidae, Pentatomini Review and cladistic analysis of Serdia Stål (Heteroptera, Pentatomidae, Pentatomini

Directory of Open Access Journals (Sweden)

Nora Denise Fortes de Fortes

2005-09-01

Full Text Available Treze espécies são hoje incluídas no gênero: S. apicicornis, Stål, 1860; S. beckerae Thomas & Rolston, 1985; S. calligera Stål, 1860; S. concolor Ruckes, 1958; S. costalis Ruckes, 1958; S. delphis Thomas & Rolston, 1985; S. inspersipes Stål, 1860; S. lobata Thomas & Rolston, 1985; S. rotundicornis Becker, 1967 e S. ruckesi Thomas & Rolston, 1985. Cinco novas espécies são descritas: S. indistincta sp. nov (Irai, Rio Grande do Sul, S. bicolor sp. nov (Ponta Grossa, Paraná, S. maculata sp. nov (Itatiaia, Rio de Janeiro, S. máxima sp. nov (Imbituba, Santa Catarina e S. robusta sp. nov (Itatiaia, Rio de Janeiro do Brasil. A análise cladística foi realizada usando 40 caracteres e 21 táxons. O gênero Tibilis Stål, 1860; Neotibilis Grazia & Barcellos, 1994 e Similliserdia Fortes & Grazia, 1998 foram usados como grupo-externo. A monofilia de Serdia foi sustentada por 3 sinapomorfias: ápice do escutelo com margens enegrecidas, machos com a parede da taça genital espessada com processos em aba, fêmeas com o espessamento da íntima vaginal situado na metade posterior das gonapófises 9 e projetando-se ventralmente. O subgênero Brasiliicola Kirkaldy, 1909 é considerado sinônimo junior de Serdia. São fornecidas ilustrações, mapas de distribuição geográfica e chave para as espécies.Thirteen species are presently included in the genus: S. apicicornis Stål, 1860; S. beckerae Thomas & Rolston, 1985; S. bihamulata, Thomas & Rolston, 1985; S. calligera Stål, 1860; S. concolor Ruckes, 1958; S. costalis Ruckes, 1958; S. delphis Thomas & Rolston, 1985; S. inspersipes Stål, 1860; S. limbatipennis Stål, 1860; S. lobata Thomas & Rolston, 1985; S. quadridens Thomas & Rolston, 1985; S. rotundicornis Becker, 1967, and S. ruckesi Thomas & Rolston, 1985. Five new species are described: S.indistincta sp. nov. (Iraí; Rio Grande do Sul, S. bicolor sp. nov. (Ponta Grossa; Paraná S. maculata sp. nov. (Itatiaia; Rio de Janeiro S. maxima sp. nov

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.

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...

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.

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.

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...

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...

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.

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.

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.)

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

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...

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

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...

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

DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

International Nuclear Information System (INIS)

CUNNANE, J.

2004-01-01

Several hundred distinct types of DOE-owned spent nuclear fuel (DSNF) may potentially be disposed in the Yucca Mountain repository. These fuel types represent many more types than can be viably individually examined for their effect on the Total System Performance Assessment for the License Application (TSPA-LA). Additionally, for most of these fuel types, there is no known direct experimental test data for the degradation and dissolution of the waste form in repository groundwaters. The approach used in the TSPA-LA model is, therefore, to assess available information on each of 11 groups of DSNF, and to identify a model that can be used in the TSPA-LA model without differentiating between individual codisposal waste packages containing different DSNF types. The purpose of this report is to examine the available data and information concerning the dissolution kinetics of DSNF matrices for the purpose of abstracting a degradation model suitable for use in describing degradation of the DSNF inventory in the Total System Performance Assessment for the License Application. The data and information and associated degradation models were examined for the following types of DSNF: Group 1--Naval spent nuclear fuel; Group 2--Plutonium/uranium alloy (Fermi 1 SNF); Group 3--Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF); Group 4--Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF); Group 5--Thorium/uranium carbide (Fort St. Vrain SNF); Group 6--Thorium/uranium oxide (Shippingport light water breeder reactor SNF); Group 7--Uranium metal (N Reactor SNF); Group 8--Uranium oxide (Three Mile Island-2 core debris); Group 9--Aluminum-based SNF (Foreign Research Reactor SNF); Group 10--Miscellaneous Fuel; and Group 11--Uranium-zirconium hydride (Training Research Isotopes-General Atomics SNF). The analyses contained in this document provide an ''upper-limit'' (i

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...

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...

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

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....

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.

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...

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

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.

Potential of small nuclear reactors for future clean and safe energy sources

International Nuclear Information System (INIS)

Sekimoto, H.

1992-01-01

To cope with the various kinds of energy demands expected in the 21st century, it is necessary to explore the potential of small nuclear reactors and to find a way of promoting their introduction to society. The main goal of current research activities is 'the constitution of the self-consistent nuclear energy system'. These activities can be understood by realizing that the nuclear community is facing a turning point for its survival in the 21st century. Self-consistency can be manifested by investigating and developing the potential advantages of the nuclear fission reaction and lessening the potential disadvantages. The contributions in this volume discuss concepts of small reactors, applications of small reactors, and consistency with conventional energy supply systems

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

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 ...

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)

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.

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)

Development of antibiotic resistance genes in microbial communities during long-term operation of anaerobic reactors in the treatment of pharmaceutical wastewater.

Science.gov (United States)

Aydin, Sevcan; Ince, Bahar; Ince, Orhan

2015-10-15

Biological treatment processes offer the ideal conditions in which a high diversity of microorganisms can grow and develop. The wastewater produced during these processes is contaminated with antibiotics and, as such, they provide the ideal setting for the acquisition and proliferation of antibiotic resistance genes (ARGs). This research investigated the occurrence and variation in the ARGs found during the one-year operation of the anaerobic sequencing batch reactors (SBRs) used to treat pharmaceutical wastewater that contained combinations of sulfamethoxazole-tetracycline-erythromycin (STE) and sulfamethoxazole-tetracycline (ST). The existence of eighteen ARGs encoding resistance to sulfamethoxazole (sul1, sul2, sul3), erythromycin (ermA, ermF, ermB, msrA, ereA), tetracycline (tetA, tetB, tetC, tetD, tetE, tetM, tetS, tetQ, tetW, tetX) and class Ι integron gene (intΙ 1) in the STE and ST reactors was investigated by quantitative real-time PCR. Due to the limited availability of primers to detect ARGs, Illumina sequencing was also performed on the sludge and effluent of the STE and ST reactors. Although there was good reactor performance in the SBRs, which corresponds to min 80% COD removal efficiency, tetA, tetB, sul1, sul2 and ermB genes were among those ARGs detected in the effluent from STE and ST reactors. A comparison of the ARGs acquired from the STE and ST reactors revealed that the effluent from the STE reactor had a higher number of ARGs than that from the ST reactor; this could be due to the synergistic effects of erythromycin. According to the expression of genes results, microorganisms achieve tetracycline and erythromycin resistance through a combination of three mechanisms: efflux pumping protein, modification of the antibiotic target and modifying enzymes. There was also a significant association between the presence of the class 1 integron and sulfamethoxazole resistance genes. Copyright © 2015 Elsevier Ltd. All rights reserved.

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....

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...

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

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…

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

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

Study on conceptual design system of tritium production fusion reactor

International Nuclear Information System (INIS)

He Kaihui

2004-11-01

Conceptual design of an advanced tritium production reactor based on spherical torus, which is intermediate application of fusion energy, was presented. Different from traditional tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST were used to minimize tritium leakage and to maximize tritium breeding ratio with arrangement of tritium production blankets as possible as it can within vacuum vessel in order to produce 1 kg excess tritium except self-sufficient plasma core, corresponding plant availability 40% or more. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR was presented. Besides systematical analyses; design risk, uncertainty and backup are introduced generally for the backgrounds of next detailed conceptual design. (author)

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...

A study on conceptual design of tritium production fusion reactor based on spherical torus

International Nuclear Information System (INIS)

He Kaihui; Huang Jinhua

2003-01-01

Conceptual design of an advanced tritium production reactor based on spherical torus (ST), which is an intermediate application of fusion energy, is presented. Different from traditional Tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST are used to minimize tritium leakage and to maximize tritium breeding ratio with arrangement of tritium production blankets as possible as it can do within vacuum vessel in order to produce certain amount of excess tritium except self-sufficient plasma core, corresponding plant availability 40% or more. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR is presented. Based on systematical analysis, design risk, uncertainty and backup are introduced generally for the backgrounds of next detailed conceptual design. (authors)

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.

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.

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.

Construction schedule management of China Experimental Fast Reactor

International Nuclear Information System (INIS)

Wang Yue

2012-01-01

China Experimental Fast Reactor (CEFR) in the first Fast Reactor in China, which is one of large project of the National High Technology Research and Development Program ('863' Program). On 21 st July 2011, CEFR had succeeded to connect to power grid, the target of construction had come true. To a large item, schedule management is one of the most important management, this paper a overall discussion about CEFR item. It has proved that the management of CEFR project is scientific, normative and high-efficiency, it will be valuable for lager Fast Reactor item and designers in interrelated field. (author)

Application of real-time PCR to determination of combined effect of antibiotics on Bacteria, Methanogenic Archaea, Archaea in anaerobic sequencing batch reactors.

Science.gov (United States)

Aydin, Sevcan; Ince, Bahar; Ince, Orhan

2015-06-01

This study evaluated the long-term effects of erythromycin-tetracycline-sulfamethoxazole (ETS) and sulfamethoxazole-tetracycline (ST) antibiotic combinations on the microbial community and examined the ways in which these antimicrobials impact the performance of anaerobic reactors. Quantitative real-time PCR was used to determine the effect that different antibiotic combinations had on the total and active Bacteria, Archae and Methanogenic Archae. Three primer sets that targeted metabolic genes encoding formylterahydrofolate synthetase, methyl-coenzyme M reductase and acetyl-coA synthetase were also used to determine the inhibition level on the mRNA expression of the homoacetogens, methanogens and specifically acetoclastic methanogens, respectively. These microorganisms play a vital role in the anaerobic degradation of organic waste and targeting these gene expressions offers operators or someone at a treatment plant the potential to control and the improve the anaerobic system. The results of the investigation revealed that acetogens have a competitive advantage over Archaea in the presence of ETS and ST combinations. Although the efficiency with which methane production takes place and the quantification of microbial populations in both the ETS and ST reactors decreased as antibiotic concentrations increased, the ETS batch reactor performed better than the ST batch reactor. According to the expression of genes results, the syntrophic interaction of acetogens and methanogens is critical to the performance of the ETS and ST reactors. Failure to maintain the stability of these microorganisms resulted in a decrease in the performance and stability of the anaerobic reactors. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

Department of Energy: Nuclear S&T workforce development programs

International Nuclear Information System (INIS)

Bingham, Michelle; Bala, Marsha; Beierschmitt, Kelly; Steele, Carolyn; Sattelberger, Alfred P.; Bruozas, Meridith A.

2016-01-01

The U.S. Department of Energy (DOE) national laboratories use their expertise in nuclear science and technology (S&T) to support a robust national nuclear S&T enterprise from the ground up. Traditional academic programs do not provide all the elements necessary to develop this expertise, so the DOE has initiated a number of supplemental programs to develop and support the nuclear S&T workforce pipeline. This document catalogs existing workforce development programs that are supported by a number of DOE offices (such as the Offices of Nuclear Energy, Science, Energy Efficiency, and Environmental Management), and by the National Nuclear Security Administration (NNSA) and the Naval Reactor Program. Workforce development programs in nuclear S&T administered through the Department of Homeland Security, the Nuclear Regulatory Commission, and the Department of Defense are also included. The information about these programs, which is cataloged below, is drawn from the program websites. Some programs, such as the Minority Serving Institutes Partnership Programs (MSIPPs) are available through more than one DOE office, so they appear in more than one section of this document.

Conceptual design of tritium production fusion reactor based on spherical torus

International Nuclear Information System (INIS)

He Kaihui; Huang Jinhua

2003-01-01

Conceptual design of an advanced tritium production fusion reactor based on spherical torus, which is intermediate application of fusion energy, was presented in this paper. Differing from the traditional tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST were used to minimize tritium leakage and maximize tritium breeding ratio with arrangement of tritium production blankets within vacuum vessel as possible in order to produce 1 kg excess tritium except need of self-sufficient plasma core with 40% or more corresponding plant availability. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR was presented, providing the backgrounds and reference for next detailed conceptual design

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.

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)

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.)

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)

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.

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

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

Sustainable Nuclear Energy for the 21st Century

International Nuclear Information System (INIS)

2010-09-01

Concerns over energy resource availability, energy security and climate change suggest an important role for nuclear power in supplying sustainable energy in the 21st century. The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was initiated in 2000 by a resolution of the IAEA General Conference to help ensure that nuclear energy is available to contribute to meeting global energy needs of the 21st century in a sustainable manner. It is a mechanism for IAEA Member States that have joined the project as INPRO members to collaborate on topics of joint interest. By 2010, INPRO membership had grown to 30 countries and the European Commission. The results of INPRO's activities, however, are made available to all IAEA Member States

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)

Det foto-eliciterede interview

DEFF Research Database (Denmark)

Rasmussen, Kim

2017-01-01

Det foto-eliciterede interview fremkalder informationer og fortællinger ud af fotografier, og støtter børn i at ytre sig.......Det foto-eliciterede interview fremkalder informationer og fortællinger ud af fotografier, og støtter børn i at ytre sig....

The ARIES tokamak fusion reactor study

International Nuclear Information System (INIS)

Bartlit, J.R.; Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Beecraft, W.R.; Hogan, J.T.; Peng, Y.K.M.; Reid, R.L.; Strickler, D.J.; Whitson, J.C.; Blanchard, J.P.; Emmert, G.A.; Santarius, J.F.; Sviatoslavsky, I.N.; Wittenberg, L.J.

1989-01-01

The ARIES study is a community effort to develop several visions of the tokamak as fusion power reactors. The aims are to determine their potential economics, safety, and environmental features and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak in 2nd stability regime and employs both potential advances in the physics and expected advances in technology and engineering; and ARIES-III is a conceptual D 3 He reactor. This paper focuses on the ARIES-I design. Parametric systems studies show that the optimum 1st stability tokamak has relatively low plasma current (∼ 12 MA), high plasma aspect ratio (∼ 4-6), and high magnetic field (∼ 24 T at the coil). ARIES-I is 1,000 MWe (net) reactor with a plasma major radius of 6.5 m, a minor radius of 1.4 m, a neutron wall loading of about 2.8 MW/m 2 , and a mass power density of about 90 kWe/ton. The ARIES-I reactor operates at steady state using ICRF fast waves to drive current in the plasma core and lower-hybrid waves for edge-plasma current drive. The current-drive system supplements a significant (∼ 57%) bootstrap current contribution. The impurity control system is based on high-recycling poloidal divertors. Because of the high field and large Lorentz forces in the toroidal-field magnets, innovative approaches with high-strength materials and support structures are used. 24 refs., 4 figs., 1 tab

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....

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.

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

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

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 ...

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...

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...

DSNF AND OTHER WASTE FORM DEGRADATION ABSTRACTION

Energy Technology Data Exchange (ETDEWEB)

J. CUNNANE

2004-11-19

Several hundred distinct types of DOE-owned spent nuclear fuel (DSNF) may potentially be disposed in the Yucca Mountain repository. These fuel types represent many more types than can be viably individually examined for their effect on the Total System Performance Assessment for the License Application (TSPA-LA). Additionally, for most of these fuel types, there is no known direct experimental test data for the degradation and dissolution of the waste form in repository groundwaters. The approach used in the TSPA-LA model is, therefore, to assess available information on each of 11 groups of DSNF, and to identify a model that can be used in the TSPA-LA model without differentiating between individual codisposal waste packages containing different DSNF types. The purpose of this report is to examine the available data and information concerning the dissolution kinetics of DSNF matrices for the purpose of abstracting a degradation model suitable for use in describing degradation of the DSNF inventory in the Total System Performance Assessment for the License Application. The data and information and associated degradation models were examined for the following types of DSNF: Group 1--Naval spent nuclear fuel; Group 2--Plutonium/uranium alloy (Fermi 1 SNF); Group 3--Plutonium/uranium carbide (Fast Flux Test Facility-Test Fuel Assembly SNF); Group 4--Mixed oxide and plutonium oxide (Fast Flux Test Facility-Demonstration Fuel Assembly/Fast Flux Test Facility-Test Demonstration Fuel Assembly SNF); Group 5--Thorium/uranium carbide (Fort St. Vrain SNF); Group 6--Thorium/uranium oxide (Shippingport light water breeder reactor SNF); Group 7--Uranium metal (N Reactor SNF); Group 8--Uranium oxide (Three Mile Island-2 core debris); Group 9--Aluminum-based SNF (Foreign Research Reactor SNF); Group 10--Miscellaneous Fuel; and Group 11--Uranium-zirconium hydride (Training Research Isotopes-General Atomics SNF). The analyses contained in this document provide an &apos

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)

Status of Dalat research reactor and progress of new reactor plan in Vietnam

International Nuclear Information System (INIS)

Dien, Nguyen Nhi; Vien, Luong Ba

2005-01-01

The Dalat Nuclear Research Reactor (DNRR) is a 500-kW pool-type reactor loaded with the Soviet WWR-M2 Fuel Assemblies (FA), moderated and cooled by light water. The reactor was reconstructed from the USA 250-kW TRIGA Mark-II reactor built in early 1960s. The first criticality of the renovated reactor was achieved on 1 st November 1983, and then on 20 March 1984 the reactor was officially inaugurated and its activities restarted. During the last twenty years, the DNRR has played an important role as a large national research facility to implement researches and applications, and its utilization has been broadened in various fields of human life. However, due to the limitation of the neutron flux and power level, the out-of date design of the experimental facilities and the ageing of the reactor facilities, it cannot meet the increasing user's demands even in the existing utilization areas. In addition, the utilization demands of the Research Reactor (RR) will be increased along with the development of the nation's economy growth. In this aspect, it is necessary to have in Vietnam a new high performance multipurpose RR with a sufficient neutron flux and power level. According to the last draft of a national strategy for atomic energy development submitted to the Government for consideration and approval, it is expected that a new high power RR would be put into operation before 2020. The operation and utilization status of the DNRR is presented and some preliminary results of the national research project on new reactor plan for Vietnam are discussed in this paper

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.

77 FR 22475 - Standard Instrument Approach Procedures, and Takeoff Minimums and Obstacle Departure Procedures...

Science.gov (United States)

2012-04-16

..., Orig, CANCELLED Fort Huachuca Sierra Vista, AZ, Sierra Vista Muni-Libby AAF, RADAR-1, Orig Fort Huachuca Sierra Vista, AZ, Sierra Vista Muni-Libby AAF, RADAR-2, Orig Lake Havasu City, AZ, Lake Havasu... Opelousas, LA, St Landry Parish-Ahart Field, NDB RWY 18, Amdt 3 Opelousas, LA, St Landry Parish-Ahart Field...

Phenolic Wastewater Treatment using Activated Carbon in a Three Phase Fluidized-Bed Reactor

Directory of Open Access Journals (Sweden)

Pornsiri Tongprem

2009-11-01

Full Text Available Phenolic wastewater treatment was investigated using activated carbon in a lab scale three phase fluidized-bed reactor. The reactor with effective volume of 272 ml, 300 mm in height and 40 mm in diameter was made from transparent acrylic that allowed to observe the phenomena occurring inside. Phenol 10 mg/l and air were used as representative agents that were continuously fed to the reactor at a constant flow rate of 1 and 2 l/min with co-current and up-flow, respectively. Comparison of the phenolic adsorption under five different conditions: (a fresh Acs, (b 1st reused Acs, (c fresh Fe/Acs, (d 1st reused Fe/Acs, and (e 2nd reused Fe/Acs, have been carried out. The phenolic wastewater was re-circulated through the reactor and its concentration was measured with respect to time. The experimental adsorption results revealed that both fresh Acs and Fe/Acs gave the better results than reused Acs and reused Fe/Acs, respectively. The adsorption in all cases of Acs and Fe/Acs would follow Pseudo-second order kinetic.

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...

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).

Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

Energy Technology Data Exchange (ETDEWEB)

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01

Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle

The feasibility study on commercialized fast reactor cycle system

International Nuclear Information System (INIS)

Noda, Hiroshi

2002-01-01

The feasibility study on commercialized Fast Reactor cycle system (FS) has been carried out by a joint team with the participation of all parties concerned in Japan since July, 1999. It aims to clarify various perspectives for commercialized fast reactor cycle system and also suggest development strategies to diverse social needs in the 21 st century. The FS consists of several phases. The phase 1 has progressed as planned and the highly feasible candidate concepts with innovative technologies have been screened out among a wide variety of concepts. During the phase 2, approximately five years after the phase 1, the in-depth design studies and engineering scale tests of key technologies are being conducted to verify and validate the feasibility of screened candidate concepts. At the end of the phase 2, a few promising concepts will be selected with their R and D tasks. The paper describes the results of the phase 1, the activities of the phase 2 and the new concept related to the fast reactor fuel cycle focusing on the reduction in environmental burden, which is one of key factors to sustain the nuclear power generation in the 21 st century

Review of fast reactor activities in India (1984)

International Nuclear Information System (INIS)

Paranjpe, S.R.

1986-01-01

During the year a number of reviews and construction activities have been practically completed as required for the 1st criticality of FBTR. The reactor is expected to become critical by the middle of 1985. The design studies for 500 MWe prototype fast breeder reactor (PFBR) have been continued. Due to preoccupation with the completion of construction of FBTR, the limited effort has been focussed on the design of key components like the sodium pumps, drivers for sodium pumps, control rod drive mechanism and steam generators. The main programs, which are a continuing activity in RRC, are discussed in this report. They are: reactor physics, radio-chemistry, metallurgy, reprocessing and safety research

Fusion reactor design: On the road to commercialization

International Nuclear Information System (INIS)

Kulcinski, G.L.

1984-01-01

The worldwide effort in fusion is now approximately 2 billion dollars per year and over 12 billion dollars has been invested since 1951 in developing this energy source for the 21st century. A vital component of the past efforts in fusion research has been the conceptual design activities performed by scientists and engineers around the world. Almost 80 such major designs of Tokamak, Mirror, Laser and Ion Beam Reactors have been published and this article discusses how recent conceptual designs have afftected our perception of future fusion reactor performance. (orig.) [de

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.

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...

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...

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)

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

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...

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

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...

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...

Fortællinger om sorg og tab – når det personlige bliver socialt?

DEFF Research Database (Denmark)

Christensen, Dorthe Refslund; Sandvik, Kjetil

2016-01-01

Vi fortæller om døden, om det at miste og føle sorg. Kulturhistorisk er litteratur, teater og malerkunst scener for netop dette emne. Som socialt fænomen ser vi dog ikke i samme omfang fortællinger om død, tab og sorg, eftersom emnet typisk har været anskuet som et privat anliggende. Bestemte...

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 ...

Controls Over Materiel Procured for Direct Vendor Delivery

Science.gov (United States)

1995-02-10

National Guard, Company D, 560th Engineer Battalion, Bainbridge, GA Army National Guard, Company E, 121st Infantry Battalion, Tifton , GA Joint...Command, Fort Monmouth, NJ United States Army Forces Command, Atlanta, GA United States Army Materiel Command, Alexandria, VA United States Army...Fort Gillem, GA Headquarters, Fort Lee, Petersburg, VA Headquarters, Fort Riley, KS Headquarters, National Guard Bureau, Washington, DC Headquarters

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.

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

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.

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

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...

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...

Raw materials for reflector graphite (for reactors)

International Nuclear Information System (INIS)

Wilhelmi, G.; Mindermann, D.

1992-01-01

The manufacturing concept for the core components of German high temperature reactor (HTR) types of graphite was previously entirely directed to the use of German tar coke (St coke). As the plants for producing this material no longer complied technically with the current environmental protection requirements, one had to assume that they would soon be shut down. To prevent bottlenecks in the erection of future HTR plants, alternative cokes produced by modern processes by Japanese manufacturers were checked for their suitability for the manufacture of reactor graphite. This report describes the investigations carried out on these materials from the safe delayed coking process. The project work, apart from analysis of the main data of the candidate coke considered, included the processing of the raw materials into directly and secondarily extruded graphite rods on the laboratory scale, including characterisation. As the results show, the material data achieved with the previous raw material can be reproduced with Japanese St coke. The tar coke LPC-A from the Nippon Steel Chemical Co., Ltd was decided on as the new standard coke for manufacturing reflector graphite. (orig.) With 15 tabs., 2 figs [de

La fouille du fort Saint-Georges à Chinon (Indre-et-Loire. Premiers résultats The excavation of fort Saint-Georges at Chinon (Indre-et-Loire. First results

Directory of Open Access Journals (Sweden)

Bruno Dufaÿ

2006-05-01

Full Text Available Cette note présente les premiers résultats des fouilles menées en 2003 et 2004 sur la quasi-totalité du fort Saint-Georges à Chinon (Indre-et-Loire. Celui-ci est l’un des trois éléments de la forteresse médiévale qui domine la ville. La fouille a permis de préciser la fonction du fort, construit dans la deuxième moitié du XIIe s., à l’époque où Chinon est le centre administratif des possessions continentales des Plantagenêt, rois d’Angleterre. Du point de vue militaire, il formait une fortification avancée, protégeant le château principal, selon une structure que Richard Cœur de Lion appliquera au Château Gaillard. À l’intérieur, de vastes bâtiments constituaient des logis, conçus peut-être au départ pour héberger la chancellerie royale.This article presents the first results of the excavations undertaken in 2003 and 2004 over almost all of the Fort Saint-Georges at Chinon (Indre-et-Loire, one of three elements of the medieval fortress which dominates the town. The excavation enabled us to clarify the function of the fort, built in the 2nd half of the 12th century at a time when Chinon was the administrative centre of the continental possesions of the Plantagenet King of England. From a military point of view, it formed an advanced fortification protecting the main castle, within a structure that Richard the Lionheart would apply to the Chayeau Gaillard. Inside, some vast buildings made up the dwellings, designed perhaps initially to house the royal chanceller.

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

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)

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.

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...

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...

Bioequivalence of fixed-dose combination Myrin®-P Forte and reference drugs in loose combination.

Science.gov (United States)

Wang, H F; Wang, R; O'Gorman, M; Crownover, P; Naqvi, A; Jafri, I

2013-12-01

Myrin®-P Forte is a fixed-dose combination (FDC) tablet containing rifampicin (RMP, 150 mg), isoniazid (INH, 75 mg), ethambutol (EMB) hydrochloride (275 mg) and pyrazinamide (PZA, 400 mg) developed for the treatment of tuberculosis (TB). This study was conducted at a single centre--the Pfizer Clinical Research Unit in Singapore. To demonstrate the bioequivalence of each drug component of the Myrin-P Forte FDC and the individual product in loose combination. In a randomized, open-label, single-dose, two-way, crossover study, subjects received single doses of Myrin-P Forte or four individual products under fasting conditions in a crossover fashion with at least 7 days washout between doses. The primary measures for comparison were peak plasma concentration (C(max)) and the area under plasma concentration-time curve (AUC). Of 36 subjects enrolled, 35 completed the study. The adjusted geometric mean ratios and 90% confidence intervals for C(max) and AUC values were completely contained within bioequivalence limits (80%, 125%) for all four drugs in both formulations. Both treatments were generally well tolerated in the study. The Myrin-P Forte FDC tablet formulation is bioequivalent to the four single-drug references for RMP, INH, EMB hydrochloride and PZA at equivalent doses.

Pakistan research reactor and its utilization

International Nuclear Information System (INIS)

Iqbal Hussain Qureshi; Naeem Ahmad Khan.

1983-01-01

The 5 MW enriched uranium fuelled, light water moderated and cooled Pakistan Research reactor became critical on 21st December, 1965 and was taken to full power on 22nd June, 1966. Since then is has been operated for about 23000 hours till 30th June, 1983 without any major break down. It has been used for the studies of neutron cross-sections, nuclear structure, fission physics, structure of material, radiation damage in crystals and semiconductors, studies of geological, biological and environmental samples by neutron activation techniques, radioisotope production, neutron radiography and for training of scientists, engineers and technicians. In the paper we have described briefly the facility of Pakistan Research Reactor and the major work carried around it during the last decade. (author)

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.

Hydrogeochemical cycling and chemical denudation in the Fort River Watershed, central Massachusetts: An appraisal of mass-balance studies

Science.gov (United States)

Yuretich, Richard F.; Batchelder, Gail L.

1988-01-01

The Fort River watershed in central Massachusetts receives precipitation with a composition similar to that in Hubbard Brook (New Hampshire), yet the average stream water chemistry is substantially different, showing higher pH and TDS. This is largely a function of bedrock and surficial geology, and chemical differences among small streams within the Fort River watershed are apparently controlled by the composition and thickness of the prevailing surficial cover. The surficial deposits determine groundwater and surface water flow paths, thereby affecting the resultant contact time with mineral matter and the chemistry of the runoff. Despite the rural setting, over 95% of the annual sodium and chloride in the streams comes from road salt; after correcting for this factor, cation denudation rates are about equal to those at Hubbard Brook. However, silica removal is occurring at a rate more than 30% greater in the Fort River. When climatic conditions in Hubbard Brook and Fort River are normalized, weathering rates appear consistently higher in the Fort River, reflecting differences in weathering processes (i.e., cation exchange and silicate breakdown) and hydrogeology. Because of uncertainties in mechanisms of cation removal from watersheds, the silica denudation rate may be a better index of weathering intensity.

Energy efficiency campaign for residential housing at the Fort Lewis army installation

Energy Technology Data Exchange (ETDEWEB)

AH McMakin; RE Lundgren; EL Malone

2000-02-23

In FY1999, Pacific Northwest National Laboratory conducted an energy efficiency campaign for residential housing at the Fort Lewis Army Installation near Tacoma, Washington. Preliminary weather-corrected calculations show energy savings of 10{percent} from FY98 for energy use in family housing. This exceeded the project's goal of 3{percent}. The work was funded by the U.S. DOEs Federal Energy Management Program (FEMP), Office of Energy Efficiency and Renewable Energy. The project adapted FEMP's national ``You Have the Power Campaign'' at the local level, tailoring it to the military culture. The applied research project was designed to demonstrate the feasibility of tailored, research-based strategies to promote energy conservation in military family housing. In contrast to many energy efficiency efforts, the campaign focused entirely on actions residents could take in their own homes, as opposed to technology or housing upgrades. Behavioral change was targeted because residents do not pay their own utility bills; thus other motivations must drive personal energy conservation. This campaign augments ongoing energy savings from housing upgrades carried out by Fort Lewis. The campaign ran from September 1998 through August 1999. The campaign strategy was developed based on findings from previous research and on input from residents and officials at Fort Lewis. Energy use, corrected to account for weather differences, was compared with the previous year's use. Survey responses from 377 of Fort Lewis residents of occupied housing showed that the campaign was moderately effective in promoting behavior change. Of those who were aware of the campaign, almost all said they were now doing one or more energy-efficient things that they had not done before. Most people were motivated by the desire to do the right thing and to set a good example for their children. They were less motivated by other factors.

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.

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

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

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.

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...

The Coast Artillery Journal. Volume 80, Number 6, November-December 1937

Science.gov (United States)

1937-12-01

cluded in the trip were the posts of Fort Kamehameha and Fort DeRussy, where the big guns were given the once- over by the Congressmen. They all seemed...November-Decembt:,; batteries have been taking places at Fort Kamehameha . tember 13th and 20th, the following batteries fired their These are additional...55th COAST ARTIL- LERY 61st COAST ARTIL- Fort Kamehameha . T. H. LERY ~IA.JOR Fort Sheridan, III. Phillip’, \\\\’. S. CAPTAIX~ Dod~~. F. B .. ,Jr

USASOC Injury Prevention/Performance Optimization Musculoskeletal Screening Initiative

Science.gov (United States)

2016-10-29

PERFORMING ORGANIZATION REPORT NUMBER er, 3860 South Water St Pittsburgh PA 15203 USASOC 1105 El Salvador St Building E-3323 Fort Bragg NC 28310 9...society beyond science and technology? • improving public knowledge , attitudes, skills, and abilities; • changing behavior, practices, decision making...unit at Fort Bragg to participate in Phase 3 Aim 1, possibly the Special Warfare Center and School (SWCS) candidates for Special Forces. This proposed

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.

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

Impact of human-associated Escherichia coli clonal groups in Antarctic pinnipeds: presence of ST73, ST95, ST141 and ST131.

Science.gov (United States)

Mora, Azucena; García-Peña, Francisco Javier; Alonso, María Pilar; Pedraza-Diaz, Susana; Ortega-Mora, Luis Miguel; Garcia-Parraga, Daniel; López, Cecilia; Viso, Susana; Dahbi, Ghizlane; Marzoa, Juan; Sergeant, Martin J; García, Vanesa; Blanco, Jorge

2018-03-16

There is growing concern about the spreading of human microorganisms in relatively untouched ecosystems such as the Antarctic region. For this reason, three pinniped species (Leptonychotes weddellii, Mirounga leonina and Arctocephalus gazella) from the west coast of the Antartic Peninsula were analysed for the presence of Escherichia spp. with the recovery of 158 E. coli and three E. albertii isolates. From those, 23 harboured different eae variants (α1, β1, β2, ε1, θ1, κ, ο), including a bfpA-positive isolate (O49:H10-A-ST206, eae-k) classified as typical enteropathogenic E. coli. Noteworthy, 62 of the 158 E. coli isolates (39.2%) exhibited the ExPEC status and 27 (17.1%) belonged to sequence types (ST) frequently occurring among urinary/bacteremia ExPEC clones: ST12, ST73, ST95, ST131 and ST141. We found similarities >85% within the PFGE-macrorrestriction profiles of pinniped and human clinic O2:H6-B2-ST141 and O16:H5/O25b:H4-B2-ST131 isolates. The in silico analysis of ST131 Cplx genomes from the three pinnipeds (five O25:H4-ST131/PST43-fimH22-virotype D; one O16:H5-ST131/PST506-fimH41; one O25:H4-ST6252/PST9-fimH22-virotype D1) identified IncF and IncI1 plasmids and revealed high core-genome similarities between pinniped and human isolates (H22 and H41 subclones). This is the first study to demonstrate the worrisome presence of human-associated E. coli clonal groups, including ST131, in Antarctic pinnipeds.

Fort Lewis natural gas and fuel oil energy baseline and efficiency resource assessment

International Nuclear Information System (INIS)

Brodrick, J.R.; Daellenbach, K.K.; Parker, G.B.; Richman, E.E.; Secrest, T.J.; Shankle, S.A.

1993-02-01

The mission of the US Department of Energy (DOE) Federal Energy Management Program (FEMP) is to lead the improvement of energy efficiency and fuel flexibility within the federal sector. Through the Pacific Northwest Laboratory (PNL), FEMP is developing a fuel-neutral approach for identifying, evaluating, and acquiring all cost-effective energy projects at federal installations; this procedure is entitled the Federal Energy Decision Screening (FEDS) system. Through a cooperative program between FEMP and the Army Forces Command (FORSCOM) for providing technical assistance to FORSCOM installations, PNL has been working with the Fort Lewis Army installation to develop the FEDS procedure. The natural gas and fuel oil assessment contained in this report was preceded with an assessment of electric energy usage that was used to implement a cofunded program between Fort Lewis and Tacoma Public Utilities to improve the efficiency of the Fort's electric-energy-using systems. This report extends the assessment procedure to the systems using natural gas and fuel oil to provide a baseline of consumption and an estimate of the energy-efficiency potential that exists for these two fuel types at Fort Lewis. The baseline is essential to segment the end uses that are targets for broad-based efficiency improvement programs. The estimated fossil-fuel efficiency resources are estimates of the available quantities of conservation for natural gas, fuel oils number-sign 2 and number-sign 6, and fuel-switching opportunities by level of cost-effectiveness. The intent of the baseline and efficiency resource estimates is to identify the major efficiency resource opportunities and not to identify all possible opportunities; however, areas of additional opportunity are noted to encourage further effort

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

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...

The IAEA activities towards enhanced utilisation, sustainability and applications of research reactors

International Nuclear Information System (INIS)

Ridikas, D.; Mank, G.; Adelfang, P.; Alldred, K.; Bradley, E.E.; Goldman, I.N.; Khvan, A.; Peld, N.

2010-01-01

This paper will give a brief introduction to the programmatic structure of the Research Reactor (RR) related activities of the IAEA sub-programme 'Research Reactors', under which the project on 'Enhancement of utilization and applications of RRs' will be presented in more detail. Both recent achievements and future planed actions will be reported with the major emphasis on RR utilisation related issues, specific applications of RRs, networks and coalitions, and assistance to the Member States (MS) planning their 1st RR. (author)

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

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...

Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

International Nuclear Information System (INIS)

Yamazaki, K.; Uemura, S.; Oishi, T.; Arimoto, H.; Shoji, T.; Garcia, J.

2009-01-01

Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO 2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO 2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

The nuclear fuel cycle in the 21st century

International Nuclear Information System (INIS)

Todreas, Neil E.

2004-01-01

As we enter the 21st century and contemplate the deployment of Generation III+ machines and the development of Generation IV systems, the fuel cycle within which these reactors are to operate has become a predominant consideration. The four challenges to nuclear development of the 21st century of economics, safety, sustainability through spent fuel management and efficient fuel utilization, and proliferation resistance increasingly involve the front and back ends of the fuel cycle equally if not more than the design of the reactor which has reached a far higher level of maturity. It is tempting to accept the closed cycle with its promise of effective waste management as inevitable. The central questions, however, are the characteristics of the desired closed cycle, the relative advantages of thermal versus fast spectrum closed cycles, the character and pace of the transition to a closed cycle, and finally the most central question as to whether the closed cycle is indeed more desirable a choice than is an open cycle. The desired closed fuel cycle for the long term around which this paper is based is full actinide recycle with natural uranium feed and only fission products discharged to an ultimate waste repository. It is concluded that a major international research and development program to achieve this fuel cycle is important to pursue. However, the need to decide for the closed cycle and deploy it is not pressing for the next several decades. (author)

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

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

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)....

Study on the method of determining the sub-criticality of a reactor via the measurement of core neutron flux spatial distribution

International Nuclear Information System (INIS)

Ma Aifeng; Jiang Xiaofeng; Zhang Shaohong

2007-01-01

A new methodology based on rigorous reactor physics theory astead of the point reactor assumption was proposed to determine or monitor the sub-criticality ora reactor, especially the sub-critical reactor of ADS, via the measurement of in-core flux spatial distribution. Preliminary numerical studies on the 1st ADS sub-critical experimental facilities-Venus No.1 in China have demonstrated the feasibility of this new method. Related discussions pointed out the potential applications of the method. (authors)

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.

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...

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

Application of carbon-coated TiO2 for decomposition of methylene blue in a photocatalytic membrane reactor

International Nuclear Information System (INIS)

Mozia, Sylwia; Toyoda, Masahiro; Inagaki, Michio; Tryba, Beata; Morawski, Antoni W.

2007-01-01

An application of carbon-coated TiO 2 for decomposition of methylene blue (MB) in a photocatalytic membrane reactor (PMR), coupling photocatalysis and direct contact membrane distillation (DCMD) was investigated. Moreover, photodegradation of a model pollutant in a batch reactor without membrane distillation (MD) was also examined. Carbon-modified TiO 2 catalysts containing different amount of carbon and commercially available TiO 2 (ST-01) were used in this study. The carbon-coated catalyst prepared from a mixture of ST-01 and polyvinyl alcohol in the mass ratio of 70/30 was the most effective in degradation of MB from all of the photocatalysts applied. Photodecomposition of MB on the recovered photocatalysts was lower than on the fresh ones. The photodegradation of MB in the PMR was slower than in the batch reactor, what probably resulted from shorter time of exposure of the catalyst particles to UV irradiation. The MD process could be successfully applied for separation of photocatalyst and by-products from the feed solution

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.))

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.

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...

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

How confident is Fort McKay that industry can reclaim oil sand development

Energy Technology Data Exchange (ETDEWEB)

Fitzpatrick, C. [Fort McKay First Nations, AB (Canada)

2004-02-05

This presentation described how traditional environmental knowledge (TEK) can provide valuable information for both the reclamation design and assessment of oil sand development in Fort McKay. Conservation is valued by the Fort McKay First Nations communities who claim that current reclamation methods are too slow, and that the land is not being brought back to its original use with the uniqueness of the boreal landscape. Elders have noted that each year the water level in the Athabasca River is lower. The blowing tailings and coke dust are causing trees to dye and driving animals away. There is concern that the animals that remain may not be safe to eat. The Fort McKay First Nation community has stated that it will view reclamation as a success only when it functions with proof over many generations. The major concerns include: salt in the water draining from reclaimed areas; salt in the soils of reclaimed area; muskeg cannot be recreated; and, the issue of whether cranberry, blueberry and streambank forest areas can be recreated, along with traditional medicinal plants. Other concerns include the loss of rivers such as the Beaver Creek and Tar River, and that the water in reclaimed areas may not be suitable for animals to live in or to drink. tabs., figs.

How confident is Fort McKay that industry can reclaim oil sand development

International Nuclear Information System (INIS)

Fitzpatrick, C.

2004-01-01

This presentation described how traditional environmental knowledge (TEK) can provide valuable information for both the reclamation design and assessment of oil sand development in Fort McKay. Conservation is valued by the Fort McKay First Nations communities who claim that current reclamation methods are too slow, and that the land is not being brought back to its original use with the uniqueness of the boreal landscape. Elders have noted that each year the water level in the Athabasca River is lower. The blowing tailings and coke dust are causing trees to dye and driving animals away. There is concern that the animals that remain may not be safe to eat. The Fort McKay First Nation community has stated that it will view reclamation as a success only when it functions with proof over many generations. The major concerns include: salt in the water draining from reclaimed areas; salt in the soils of reclaimed area; muskeg cannot be recreated; and, the issue of whether cranberry, blueberry and streambank forest areas can be recreated, along with traditional medicinal plants. Other concerns include the loss of rivers such as the Beaver Creek and Tar River, and that the water in reclaimed areas may not be suitable for animals to live in or to drink. tabs., figs

The Fort McMurray Demonstration Project in Social Marketing: theory, design, and evaluation.

Science.gov (United States)

Guidotti, T L; Ford, L; Wheeler, M

2000-02-01

The Fort McMurray Demonstration Project in Social Marketing is a multifaceted program that applies the techniques of social marketing to health and safety. This paper describes the origins of the project and the principles on which it was based. VENUE: Fort McMurray, in the province of Alberta, Canada, was selected because the community had several community initiatives already underway and the project had the opportunity to demonstrate "value added." The project is distinguished from others by a model that attempts to achieve mutually reinforcing effects from social marketing in the community as a whole and from workplace safety promotion in particular. Specific interventions sponsored by the project include a media campaign on cable television, public activities in local schools, a community safety audit, and media appearance by a mascot that provides visual identity to the project, a dinosaur named "Safetysaurus." The project integrated its activities with other community initiatives. The evaluation component emphasizes outcome measures. A final evaluation based on injury rates and attitudinal surveys is underway. Baseline data from the first round of surveys have been compiled and published. In 1995, Fort McMurray became the first city in North America to be given membership in the World Health Organization's Safe Community Network.

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...

Three members of Medicago truncatula ST family are ubiquitous during development and modulated by nutritional status (MtST1) and dehydration (MtST2 and MtST3).

Science.gov (United States)

Albornos, Lucía; Martín, Ignacio; Labrador, Emilia; Dopico, Berta

2017-07-10

ShooT specific/Specific Tissue (ST) belong to a protein family of unknown function characterized by the DUF2775 domain and produced in specific taxonomic plant families, mainly Fabaceae and Asteraceae, with the Medicago truncatula ST family being the largest. The putative roles proposed for this family are cell elongation, biotic interactions, abiotic stress and N reserve. The aim of this work was to go deeper into the role of three M. truncatula ST proteins, namely ST1, ST2 and ST3. Our starting hypothesis was that each member of the family could perform a specific role, and hence, each ST gene would be subjected to a different type of regulation. The search for cis-acting regulatory elements (CREs) in silico in pST1, pST2 and pST3 promoters showed prevalence of tissue/organ specific motifs, especially root- and seed-specific ones. Light, hormone, biotic and abiotic related motifs were also present. None of these pSTs showed the same combination of CREs, or presented the same activity pattern. In general, pST activity was associated with the vascular cylinder, mainly in roots. Promoter activation was highly specific and dissimilar during reproductive development. The ST1, ST2 and ST3 transcripts accumulated in most of the organs and developmental stages analysed - decreasing with age - and expression was higher in the roots than in the aerial parts and more abundant in light-grown plants. The effect of the different treatments on transcript accumulation indicated that ST1 behaved differently from ST2 and ST3, mainly in response to several hormones and dehydration treatments (NaCl or mannitol), upon which ST1 transcript levels decreased and ST2 and ST3 levels increased. Finally, the ST1 protein was located in the cell wall whereas ST2 and ST3 were present both in the cytoplasm and in the cell wall. The ST proteins studied are ubiquitous proteins that could perform distinct/complementary roles in plant biology as they are encoded by differentially regulated genes

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

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...

Dimensions of Velopharyngeal Space following Maxillary Advancement with Le Fort I Osteotomy Compared to Zisser Segmental Osteotomy: A Cephalometric Study

Directory of Open Access Journals (Sweden)

Furkan Erol Karabekmez

2015-01-01

Full Text Available The objectives of this study are to assess the velopharyngeal dimensions using cephalometric variables of the nasopharynx and oropharynx as well as to compare the Le Fort I osteotomy technique to Zisser’s anterior maxillary osteotomy technique based on patients’ outcomes within early and late postoperative follow-ups. 15 patients with severe maxillary deficiency treated with Le Fort I osteotomy and maxillary segmental osteotomy were assessed. Preoperative, early postoperative, and late postoperative follow-up lateral cephalograms, patient histories, and operative reports are reviewed with a focus on defined cephalometric landmarks for assessing velopharyngeal space dimension and maxillary movement (measured for three different tracing points. A significant change was found between preoperative and postoperative lateral cephalometric measurements regarding the distance between the posterior nasal spine and the posterior pharyngeal wall in Le Fort I osteotomy cases. However, no significant difference was found between preoperative and postoperative measurements in maxillary segmental osteotomy cases regarding the same measurements. The velopharyngeal area calculated for the Le Fort I osteotomy group showed a significant difference between the preoperative and postoperative measurements. Le Fort I osteotomy for advancement of upper jaw increases velopharyngeal space. On the other hand, Zisser’s anterior maxillary segmental osteotomy does not alter the dimension of the velopharyngeal space significantly.

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...

Field Demonstration of Aviation Turbine Fuel MIL-T-83133C, Grade JP-8 (NATO Code F-34), at Fort Bliss, TX

Science.gov (United States)

1992-09-01

APO NY 09052 CDR US ARMY NATICK RD&E CTR DOD PROJ MGR, MOBILE ELECTRIC POWER ATTN: SATNC-US US ARMY TROOP SUPPORT COMMAND NATICK MA 01760-5020 ATUN ...US ARMY QUARTERMASTER SCHOOL ATUN : LOEA-PL (MR LeVAN) I ATTN: ATSM-CDM 1 NEW CUMBERLAND PA 17070 ATSM-PWD I FORT LEE VA 23801 PETROLEUM FIELD OFFICE...ARTILLERY CENTER US ARMY INFANTRY SCHOOL & FORT BLISS ATTN: ATSH-CD-MIS-M I ATUN : ATZC-ISL-PP 3 ATSH-CD-TSM-T 1 ATZC-ISL-MM 3 FORT BENNING GA 31905-5400

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...

Archaeological Surveys and Evaluations of Four Construction Areas in the Vicinity of Fort Jackson, Plaquemines Parish, Louisiana

Science.gov (United States)

1992-04-01

the officers’ quarters, a hospital, and an inspector’s quarters (Greene 1982:128-129). The fort itself was a regular pentagon with bastions at each...Outside of the moat another brick wall was constructed, facing a second ditch. A bridge over the second ditch led southward to a water battery whose...Archaeological Swrveys and Evaluations at Fort Jackson du Pratz, Le Page 1975 The History of Louisiana. Louisiana American Revolution Bicentennial Commission

DIGITAL PRESERVATION OF THE QUON SANG LUNG LAUNDRY BUILDING, FORT MACLEOD, ALBERTA

Directory of Open Access Journals (Sweden)

P. Dawson

2017-08-01

Full Text Available This paper describes the results of an emergency recording and archiving of a historic structure in Southern Alberta and explores the lessons learned. Digital recording of the Quon Sang Lung Laundry building in Fort Macleod, Alberta, was a joint initiative between Alberta Culture and Tourism and the University of Calgary. The Quon Sang Lung Laundry was a boomtown-style wood structure situated in the Fort Macleod Provincial Historic Area, Alberta. Built in the mid-1800s, the structure was one of the four buildings comprising Fort Macleod’s Chinatown. Its association with Chinese immigration, settlement, and emergence of Chinese-owned businesses in early twentieth-century Alberta, made the Quon Sang Lung Laundry a unique and very significant historic resource. In recent years, a condition assessment of the structure indicated that the building was not safe and that the extent of the instability could lead to a sudden collapse. In response, Alberta Culture and Tourism engaged the Departments of Anthropology and Archaeology and Geomatics Engineering from the University of Calgary, to digitally preserve the laundry building. A complete survey including the laser scanning of all the remaining elements of the original structure, was undertaken. Through digital modeling, the work guarantees that a three-dimensional representation of the building is available for future use. This includes accurate 3D renders of the exterior and interior spaces and a collection of architectural drawings comprising floor plans, sections, and elevations.

Le Fort III Distraction With Internal vs External Distractors: A Cephalometric Analysis.

Science.gov (United States)

Robertson, Kevin J; Mendez, Bernardino M; Bruce, William J; McDonnell, Brendan D; Chiodo, Michael V; Patel, Parit A

2018-05-01

This study compares the change in midface position following Le Fort III advancement using either rigid external distraction (group 1) or internal distraction (group 2). We hypothesized that, with reference to right-facing cephalometry, internal distraction would result in increased clockwise rotation and inferior displacement of the midface. Le Fort III osteotomies and standardized distraction protocols were performed on 10 cadaveric specimens per group. Right-facing lateral cephalograms were traced and compared across time points to determine change in position at points orbitale, anterior nasal spine (ANS), A-point, and angle ANB. Institutional. Twenty cadaveric head specimens. Standard subcranial Le Fort III osteotomies were performed from a coronal approach and adequately mobilized. The specified distraction mechanism was applied and advanced by 15 mm. Changes of position were calculated at various skeletal landmarks: orbitale, ANS, A-point, and ANB. Group 1 demonstrated relatively uniform x-axis advancement with minimal inferior repositioning at the A-point, ANS, and orbitale. Group 2 demonstrated marked variation in x-axis advancement among the 3 points, along with a significant inferior repositioning and clockwise rotation of the midface ( P External distraction resulted in more uniform advancement of the midface, whereas internal distraction resulted in greater clockwise rotation and inferior displacement. External distraction appears to provide increased vector control of the midface, which is important in creating a customized distraction plan based on the patient's individual occlusal and skeletal needs.

Digital Preservation of the Quon Sang Lung Laundry Building, Fort Macleod, Alberta

Science.gov (United States)

Dawson, P.; Baradaran, F.; Jahraus, A.; Rubalcava, E.; Farrokhi, A.; Robinson, C.

2017-08-01

This paper describes the results of an emergency recording and archiving of a historic structure in Southern Alberta and explores the lessons learned. Digital recording of the Quon Sang Lung Laundry building in Fort Macleod, Alberta, was a joint initiative between Alberta Culture and Tourism and the University of Calgary. The Quon Sang Lung Laundry was a boomtown-style wood structure situated in the Fort Macleod Provincial Historic Area, Alberta. Built in the mid-1800s, the structure was one of the four buildings comprising Fort Macleod's Chinatown. Its association with Chinese immigration, settlement, and emergence of Chinese-owned businesses in early twentieth-century Alberta, made the Quon Sang Lung Laundry a unique and very significant historic resource. In recent years, a condition assessment of the structure indicated that the building was not safe and that the extent of the instability could lead to a sudden collapse. In response, Alberta Culture and Tourism engaged the Departments of Anthropology and Archaeology and Geomatics Engineering from the University of Calgary, to digitally preserve the laundry building. A complete survey including the laser scanning of all the remaining elements of the original structure, was undertaken. Through digital modeling, the work guarantees that a three-dimensional representation of the building is available for future use. This includes accurate 3D renders of the exterior and interior spaces and a collection of architectural drawings comprising floor plans, sections, and elevations.

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...

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...

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...

High temperature gas-cooled reactors - once-through fuel cycle

International Nuclear Information System (INIS)

1979-03-01

The HTGR, because of a unique combination of design characteristics, is a resource-efficient and cost-effective reactor. In the HTGR, the low power density core, coated particle fuel design, and gas cooling combine to provide high neutron economy, fuel burnup and thermodynamic efficiency. The uranium resource requirements for the current MEU/Th cycle with annual refueling results in a 30-year net U 3 O 8 requirement of 4280 ST/GWe. The basic design of the HTGR refueling scheme, whereby only selected regions of the core need be accessible during each refueling, makes fuel utilization improvements through semi-annual refueling an acceptable alternative in terms of plant availability. This alternative reduces the 30-year U 3 O 8 requirement by about 9%. Additional resource utilization improvements of 10% could be realized by improved fuel management techniques. In addition to improvements achieved in reactor technology, uranium utilization can also be improved by reducing the U-235 content in the depleted uranium (tails) produced by the isotope separation facility. If the Advanced Isotope Separation Technology program, currently under development by the United States, results in a lowering of the tails assay from 0.20 w/o to 0.05 w/o the uranium feed requirement for MEU/Th cycles would be further reduced by 22%. A total improvement of 41% over the already relatively low 4280 ST/GWe net lifetime U 3 O 8 requirement would result in a 2525 ST/GWe 30-year yet U 3 O 8 requirement if all of the potential improvements were realized

Fort Hills Oil Sands Project No Net Loss Lake earthfill structure

Energy Technology Data Exchange (ETDEWEB)

Blakely, D.; Sawatsky, L. [Golder Associates Ltd., Calgary, AB (Canada); Wog, K.; Paz, S. [Alberta Environment, Edmonton, AB (Canada). Water Management Operations; Chernys, S. [Petro-Canada, Calgary, AB (Canada)

2007-07-01

The Fort Hills Oil Sands Project (FHOSP) is located north of Fort McMurray, Alberta. The Fort Hills Energy Corporation (FHEC) must compensate for fish habitat lost as a result of mine development that would disturb natural streams and lakes. FHEC planned to construct a fisheries compensation lake on the north end of its leased property, contained in part by an earthfill structure. Unlike most dam structures, the FHOSP No Net Loss Lake (NNLL) earthfill structure was planned solely for the creation of fisheries compensation habitat. Therefore, the NNLL earthfill structure must be designed with robust features that can handle any foreseeable environmental condition without failure, so that it may be accepted as a sustainable feature of the mine closure landscape. This paper discussed the design features of the NNLL earthfill structure. The paper presented information on the background of the project including regulatory criteria for the fisheries compensation habitat; fisheries compensation habitat location; and design criteria for the NNLL. The features of the NNLL earthfill structure were also discussed. In addition, the paper outlined the dam safety classification for earthfill structure and anticipated system performance. The proposed monitoring program and permanent closure plans were also discussed. It was concluded that the earthfill structure was designed with several features that would allow it to become a part of the closure landscape. These included a high width to height ratio, significant erosion protection, and an aggressive reclamation plan. These features will provide a sound basis for FHEC to apply for a reclamation certificate at the end of mine life. 3 refs., 3 tabs., 8 figs.

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

Summary progress report for fiscal year 1976 and the transition quarter describing technical assistance work for the Division of Systems Safety, U.S. Nuclear Regulatory Commission

International Nuclear Information System (INIS)

Sanders, J.P.

1977-01-01

The report reviews briefly the HTGR core analytical methods that were developed during the course of the program. The features of these analytical methods are compared with methods used to perform similar analyses, and examples of the use of these methods are cited. Included are discussions of HEATUP (a computer code for the thermal analysis of an LOFC accident in an HTGR), HEATING 5 (an IBM 360 heat-conduction code), CCCM (a coupled conduction-convection model for core thermal analysis), FLODIS (a computer model to determine the flow distribution and thermal response of the Vrain reactor), and HEXEREI 2 code development

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...

Revised Geologic Map of the Fort Garland Quadrangle, Costilla County, Colorado

Science.gov (United States)

Wallace, Alan R.; Machette, Michael N.

2008-01-01

The map area includes Fort Garland, Colo., and the surrounding area, which is primarily rural. Fort Garland was established in 1858 to protect settlers in the San Luis Valley, then part of the Territory of New Mexico. East of the town are the Garland mesas (basalt-covered tablelands), which are uplifted as horsts with the Central Sangre de Cristo fault zone. The map also includes the northern part of the Culebra graben, a deep structural basin that extends from south of San Luis (as the Sanchez graben) to near Blanca, about 8 km west of Fort Garland. The oldest rocks exposed in the map area are early Proterozic basement rocks (granites in Ikes Creek block) that occupy an intermediate structural position between the strongly uplifted Blanca Peak block and the Culebra graben. The basement rocks are overlain by Oligocene volcanic and volcaniclastic rocks of unknown origin. The volcanic rocks were buried by a thick sequence of basin-fill deposits of the Santa Fe Group as the Rio Grande rift formed about 25 million years ago. The Servilleta Basalt, a regional series of 3.7?4.8 Ma old flood basalts, was deposited within sediment, and locally provides a basis for dividing the group into upper and lower parts. Landslide deposits and colluvium that rest on sediments of the Santa Fe Group cover the steep margins of the mesas. Exposures of the sediment beneath the basalt and within the low foothills east of the Central Sangre de Cristo fault zone are comprised of siltstones, sandstones, and minor fluvial conglomerates. Most of the low ground surrounding the mesas and in the graben is covered by surficial deposits of Quaternary age. The alluvial deposits are subdivided into three Pleistocene-age units and three Holocene-age units. The oldest Pleistocene gravel (unit Qao) is preserved as isolated remnants that cap high surfaces north and east of Fort Garland. The primary geologic hazards in the map area are from earthquakes, landslides, and localized flooding. The Central

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.

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

Geologic map of the Orchard 7.5' quadrangle, Morgan County, Colorado

Science.gov (United States)

Berry, Margaret E.; Slate, Janet L.; Hanson, Paul R.; Brandt, Theodore R.

2015-01-01

The Orchard 7.5' quadrangle is located along the South Platte River corridor on the semi-arid plains of eastern Colorado, and contains surficial deposits that record alluvial, eolian, and hillslope processes that have operated through environmental changes from the Pleistocene to the present. The South Platte River, originating high in the Colorado Front Range, has played a major role in shaping the geology of the quadrangle, which is situated downstream of where the last of the major headwater tributaries (St. Vrain, Big Thompson, and Cache la Poudre) join the river. Recurrent glaciation (and deglaciation) of basin headwaters affected river discharge and sediment supply far downstream, influencing alluvium deposition and terrace formation in the Orchard quadrangle. Kiowa and Bijou Creeks, unglaciated tributaries originating east of the Front Range also have played a major role by periodically delivering large volumes of sediment to the river during flood events, which may have temporarily dammed the river. Eolian sand deposits of the Greeley (north of river) and Fort Morgan (south of river) dune fields cover much of the quadrangle and record past episodes of sand mobilization during times of drought. With the onset of irrigation during historic times, the South Platte River has changed from a broad, shallow, and sandy braided river with highly seasonal discharge to a much narrower, deeper river with braided-meandering transition morphology and more uniform discharge. Along this reach, the river has incised into Upper Cretaceous Pierre Shale, which, although buried by alluvial deposits in Orchard quadrangle, is locally exposed downstream along the South Platte River bluff near the Bijou Creek confluence, in some of the larger draws, and along Wildcat Creek.

Removal of organic substances and oxidation of ammonium nitrogen by a down-flow hanging sponge (DHS) reactor under high salinity conditions.

Science.gov (United States)

Uemura, Shigeki; Suzuki, Saori; Abe, Kenichi; Kubota, Keiichi; Yamaguchi, Takashi; Ohashi, Akiyoshi; Takemura, Yasuyuki; Harada, Hideki

2010-07-01

A down-flow hanging sponge (DHS) reactor, constructed by connecting three identical treatment units in series, was fed with highly saline artificial coke-plant wastewater containing 1400 mg L(-1) of phenol in terms of chemical oxygen demand (COD) and 500 mg-NL(-1) of ammonium nitrogen. The COD was removed by the 1st unit, achieving 92% removal at an average COD loading rate of 3.0 kg-COD m(-3)d(-1) for all units, with oxidation of ammonium nitrogen occurring primarily in the two downstream units. Microbial assays of the different units of the reactor revealed greater numbers of nitrifying bacteria in the 2nd and 3rd units than in the 1st unit, corresponding with the observed ammonium oxidation pattern of the reactor. These findings suggest that a succession of microflora was successfully established along the DHS. Copyright (c) 2010 Elsevier Ltd. All rights reserved.

The technology development for surveillance test of reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Sun Phil; Park, Day Young; Choi, Kwen Jai

1997-12-01

Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs.

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

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.

A Case Analysis of Energy Savings Performance Contract Projects and Photovoltaic Energy at Fort Bliss, El Paso, Texas

Science.gov (United States)

2006-06-01

PHOTOVOLTAIC ENERGY AND FORT BLISS CASE BACKGROUND A. PHOTOVOLTAIC ENERGY The use of photovoltaic power systems is nothing new in the Department...against the Outback MPPT charge controller . This test will be done over a one month timeframe. The Arizona Power ISG test plan is contained in...cost-benefit analysis of conventional power versus emerging photovoltaic energy for the Army’s Fort Bliss in El Paso, TX. The project will also analyze

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.

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...

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...

A Theory of Revolutionary Warfare and its Application to the Bolivian Adventure of Che Guevara

Science.gov (United States)

1973-01-01

training in Cuba Molses Nato Molses Guevara Rodriguez Julio Mendez Cano .Member of Center; later Joined the Rearguard; recrui ter Member of...Fort Bragg, North Carolina, 1965. Carol ! na. Counterinsurgency Planning Guide. ST-31-176, Fort Bragg, North Von Lsz&r, Arpad and Robert R. Kaufman

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

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

Assessment of pterygomaxillary separation in Le Fort I Osteotomy in class III patients.

Science.gov (United States)

Ueki, Koichiro; Hashiba, Yukari; Marukawa, Kohei; Okabe, Katsuhiko; Alam, Shamiul; Nakagawa, Kiyomasa; Yamamoto, Etsuhide

2009-04-01

To examine the separation of the pterygomaxillary region at the posterior nasal spine level after Le Fort I osteotomy in Class III patients. The study group consisted of 37 Japanese patients with mandibular prognathism and asymmetry, with maxillary retrognathism or asymmetry. A total of 74 sides were examined. Le Fort I osteotomy was performed without a pterygoid osteotome, with an ultrasonic curette used to remove interference at the pterygomaxillary region. Postoperative computed tomography (CT) was analyzed for all patients. The separation of the pterygomaxillary region and the location of the descending palatine artery were assessed. Although acceptable separation between the maxilla and pterygoid plates was achieved in all patients, an exact separation of the pterygomaxillary junction at the posterior nasal spine level was found in only 18 of 74 sides (24%). In 29 of 74 sides (39.2%), the separation occurred anterior to the descending palatine artery. In 29 of 74 sides (39.2%), complete separation between the maxilla and lateral and/or medial pterygoid plate was not achieved, but lower level separation of the maxilla and pterygoid plate was always complete. The maxillary segments could be moved to the postoperative ideal position in all cases. Le Fort I osteotomy without an osteotome does not always induce an exact separation at the pterygomaxillary junction at the posterior nasal spine level, but the ultrasonic bone curette can remove the interference between maxillary segment and pterygoid plates more safely.

Research, Development and Demonstration of Peak Load Reduction on Distribution Feeders Using Distributed Energy Resources for the City of Fort Collins

Energy Technology Data Exchange (ETDEWEB)

Sumner, Dennis [City of Fort Collins Utilities, CO (United States); Vosburg, Tom [City of Fort Collins Utilities, CO (United States); Brunner, Steve [Brendle Group, Fort Collins, CO (United States); Gates, Judy [Woodward, Inc., Fort Collins, CO (United States); Howard, Nathan [Spirae, Inc., Fort Collins, CO (United States); Merton, Andrew [Spirae, Inc., Fort Collins, CO (United States); Wright, Don [Spirae, Inc., Fort Collins, CO (United States); Birlingmair, Doug [Spirae, Inc., Fort Collins, CO (United States)

2015-10-01

This project titled “Research, Development and Demonstration of Peak Load Reduction on Distribution Feeders Using Distributed Energy Resources for the City of Fort Collins” evolved in response to the Department of Energy’s (DOE) Funding Opportunity Announcement (FOA) Number DE-PS26-07NT43119. Also referred to as the Fort Collins Renewable and Distributed System Integration (RDSI) Project, the effort was undertaken by a diverse group of local government, higher education and business organizations; and was driven by three overarching goals: I. Fulfill the requirements of the DOE FOA’s Area of Interest 2: Renewable and Distributed System Integration; most notably, to demonstrate the ability to reduce electric system distribution feeder peak load by 15% or more through the coordinated use of Distributed Energy Resources (DER). II. Advance the expertise, technologies and infrastructure necessary to support the long term vision of the Fort Collins Zero Energy District (FortZED) and move towards creating a zero energy district in the Fort Collins “Old Town” area. III. Further the goals of the City of Fort Collins Energy Policy, including the development of a Smart Grid-enabled distribution system in Fort Collins, expanded use of renewable energy, increased energy conservation, and peak load reduction. Through the collaborative efforts of the partner organizations, the Fort Collins RDSI project was successful in achieving all three of these goals. This report is organized into two distinct sections corresponding to the two phases of the project: • Part 1: Feeder Peak Load Reduction and the FortZED Initiative. • Part 2: Forming and Operating Utility Microgrids and Managing Load and Production Variability The original project scope addressed the Part 1 feeder peak load reduction. That work took place from 2009 through 2011 and was largely complete when the project scope was amended to include a demonstration of microgrid operations. While leveraging the

A non-conventional procedure for the 3D modeling of WWI forts

Science.gov (United States)

Nocerino, E.; Fiorillo, F.; Minto, S.; Menna, F.; Remondino, F.

2014-06-01

2014 is the hundredth anniversary of the outbreak of the First World War (WWI) - or Great War - in Europe and a number of initiatives have been planned to commemorate the tragic event. Until 1918, the Italian Trentino - Alto Adige region was under the Austro - Hungarian Empire and represented one of the most crucial and bloody war front between the Austrian and Italian territories. The region borders were constellated of military fortresses, theatre of battles between the two opposite troops. Unfortunately, most of these military buildings are now ruined and their architectures can be hardly appreciated. The paper presents the initial results of the VAST project (VAlorizzazione Storia e Territorio - Valorization of History and Landscape), that aims to digitally reconstruct the forts located on the plateaus of Luserna, Lavarone and Folgaria. An integrated methodology has been adopted to collect and employ all possible source of information in order to derive precise and photo-realistic 3D digital representations of WWI forts.

Fuel utilization potential in light water reactors with once-through fuel irradiation (AWBA Development Program)

International Nuclear Information System (INIS)

Rampolla, D.S.; Conley, G.H.; Candelore, N.R.; Cowell, G.K.; Estes, G.P.; Flanery, B.K.; Duncombe, E.; Dunyak, J.; Satterwhite, D.G.

1979-07-01

Current commercial light water reactor cores operate without recylce of fuel, on a once-through fuel cycle. To help conserve the limited nuclear fuel resources, there is interest in increasing the energy yield and, hence, fuel utilization from once-through fuel irradiation. This report evaluates the potential increase in fuel utilization of light water reactor cores operating on a once-through cycle assuming 0.2% enrichment plant tails assay. This evaluation is based on a large number of survey calculations using techniques which were verified by more detailed calculations of several core concepts. It is concluded that the maximum fuel utilization which could be achieved by practical once-through pressurized light water reactor cores with either uranium or thorium is about 17 MWYth/ST U 3 O 8 (Megawatt Years Thermal per Short Ton of U 3 O 8 ). This is about 50% higher than that of current commercial light water reactor cores. Achievement of this increased fuel utilization would require average fuel burnup beyond 50,000 MWD/MT and incorporation of the following design features to reduce parasitic losses of neutrons: reflector blankets to utilize neutrons that would otherwise leak out of the core; fuel management practices in which a smaller fraction of the core is replaced at each refueling; and neutron economic reactivity control, such as movable fuel control rather than soluble boron control. For a hypothetical situation in which all neutron leakage and parasitic losses are eliminated and fuel depletion is not limited by design considerations, a maximum fuel utilization of about 20 MWYth/ST U 3 O 8 is calculated for either uranium or thorium. It is concluded that fuel utilization for comparable reactor designs is better with uranium fuel than with thorium fuel for average fuel depletions of 30,000 to 35,000 MWD/MT which are characteristic of present light water reactor cores

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.

Digital Børnelitteratur

DEFF Research Database (Denmark)

Ferdinand, Trine

2013-01-01

Specialet drejer sig om den type digitale børnelitterære fortællinger, der skal opleves på en tabletcomputer. Disse fortællinger kan yderligere karakteriseres som interaktive og multimodale. Børnelitteratur er traditionelt blevet læst i samvær med voksne, da de voksne har været en forudsætning...... for at barnet kunne få læst den skriftlige fortælling højt, eller læsningen har foregået sammen med en lærer som en del af litteraturundervisningen i grundskolen. Dette forhold påvirkes af det digitale medie. I nyere børnelitteratur kan der ofte identificeres en dobbelt henvendelse til både barne......-og voksenlæseren i teksten, også kaldet dual adress (Wall, 1991). Der kan både identificeres en implicit barnelæser og en implicit voksenlæser i fortællingen. Rundt om digitale børnelitterære fortællinger er der indlejret et funktionelt design, der tilbyder læserne nogle særlige muligheder for at opleve, læse...

The Prospect of Neutron Scattering In the 21st Century: A Powerful Tool for Materials Research

Directory of Open Access Journals (Sweden)

E. Kartini

2007-07-01

Full Text Available Over the last 60 years research reactors (RRs have played an important role in technological and socio-economical development of mankind, such as radioisotope production for medicine, industry, research and education. Neutron scattering has been widely used for research and development in materials science. The prospect of neutron scattering as a powerful tool for materials research is increasing in the 21st century. This can be seen from the investment of several new neutron sources all over the world such as the Spallation Neutron Source (SNS in USA, the Japan Proton Accelerator Complex (JPARC in Japan, the new OPAL Reactor in Australia, and some upgrading to the existing sources at ISIS, Rutherford Appleton Laboratory, UK; Institute of Laue Langevin (ILL in Grenoble, France and Berlin Reactor, Germany. Developing countries with moderate flux research reactor have also been involved in this technique, such as India, Malaysia and Indonesia. The Siwabessy Multipurpose Reactor in Serpong, Indonesia that also produces thermal neutron has contributed to the research and development in the Asia Pacific Region. However, the international joint research among those countries plays an important role on optimizing the results.

The Prospect of Neutron Scattering in The 21st Century : A Powerful Tool For Materials Research

International Nuclear Information System (INIS)

E-Kartini

2007-01-01

Over the last 60 years research reactors (RRs) have played an important role in technological and socio-economical development of mankind, such as radioisotope production for medicine, industry, research and education. Neutron scattering has been widely used for research and development in materials science. The prospect of neutron scattering as a powerful tool for materials research is increasing in the 21 st century. This can be seen from the investment of several new neutron sources all over the world such as the Spallation Neutron Source (SNS) in USA, the Japan Proton Accelerator Complex (JPARC) in Japan, the new OPAL Reactor in Australia, and some upgrading to the existing sources at ISIS, Rutherford Appleton Laboratory, UK; Institute of Laue Langevin (ILL) in Grenoble, France and Berlin Reactor, Germany. Developing countries with moderate flux research reactor have also been involved in this technique, such as India, Malaysia and Indonesia The Siwabessy Multipurpose Reactor in Serpong, Indonesia that also produces thermal neutron has contributed to the research and development in the Asia Pacific Region. However,the international joint research among those countries plays an important role on optimizing the results. (author)

Edutainment, cultural innovation and social inclusion. Fort360, a project for cultural heritage enhancement

Directory of Open Access Journals (Sweden)

Paolo Di Pietro Martinelli

2016-12-01

Full Text Available   Fort360 project is a cultural initiative that receives the main directives of edutainment processes, trying to provide an answer to the necessity of a capillary system of information and awareness about the dismissing cultural heritage. The proposed study – carried out in the Fort Bravetta, Rome – presents a video where the educational aspect, related to the historical and architectural site contents, is strictly connected with the playful and emotional quality, resulted from a VR interaction with a panoramic video. This first case study focuses on the use of low-cost digital instrumentation and tries to improve the value of culture from the bottom, proposing an alternative way of cultural heritage enjoyment, based on participation and on interdisciplinarity of the proposed contents.

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...

An ecological response model for the Cache la Poudre River through Fort Collins

Science.gov (United States)

Shanahan, Jennifer; Baker, Daniel; Bledsoe, Brian P.; Poff, LeRoy; Merritt, David M.; Bestgen, Kevin R.; Auble, Gregor T.; Kondratieff, Boris C.; Stokes, John; Lorie, Mark; Sanderson, John

2014-01-01

The Poudre River Ecological Response Model (ERM) is a collaborative effort initiated by the City of Fort Collins and a team of nine river scientists to provide the City with a tool to improve its understanding of the past, present, and likely future conditions of the Cache la Poudre River ecosystem. The overall ecosystem condition is described through the measurement of key ecological indicators such as shape and character of the stream channel and banks, streamside plant communities and floodplain wetlands, aquatic vegetation and insects, and fishes, both coolwater trout and warmwater native species. The 13- mile-long study area of the Poudre River flows through Fort Collins, Colorado, and is located in an ecological transition zone between the upstream, cold-water, steep-gradient system in the Front Range of the Southern Rocky Mountains and the downstream, warm-water, low-gradient reach in the Colorado high plains.

The ARIES tokamak reactor study

International Nuclear Information System (INIS)

1989-10-01

The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D 3 He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions

Criticality Model

International Nuclear Information System (INIS)

Alsaed, A.

2004-01-01

The ''Disposal Criticality Analysis Methodology Topical Report'' (YMP 2003) presents the methodology for evaluating potential criticality situations in the monitored geologic repository. As stated in the referenced Topical Report, the detailed methodology for performing the disposal criticality analyses will be documented in model reports. Many of the models developed in support of the Topical Report differ from the definition of models as given in the Office of Civilian Radioactive Waste Management procedure AP-SIII.10Q, ''Models'', in that they are procedural, rather than mathematical. These model reports document the detailed methodology necessary to implement the approach presented in the Disposal Criticality Analysis Methodology Topical Report and provide calculations utilizing the methodology. Thus, the governing procedure for this type of report is AP-3.12Q, ''Design Calculations and Analyses''. The ''Criticality Model'' is of this latter type, providing a process evaluating the criticality potential of in-package and external configurations. The purpose of this analysis is to layout the process for calculating the criticality potential for various in-package and external configurations and to calculate lower-bound tolerance limit (LBTL) values and determine range of applicability (ROA) parameters. The LBTL calculations and the ROA determinations are performed using selected benchmark experiments that are applicable to various waste forms and various in-package and external configurations. The waste forms considered in this calculation are pressurized water reactor (PWR), boiling water reactor (BWR), Fast Flux Test Facility (FFTF), Training Research Isotope General Atomic (TRIGA), Enrico Fermi, Shippingport pressurized water reactor, Shippingport light water breeder reactor (LWBR), N-Reactor, Melt and Dilute, and Fort Saint Vrain Reactor spent nuclear fuel (SNF). The scope of this analysis is to document the criticality computational method. The criticality

Establishment of ambient air quality trends using historical monitoring data from Edmonton and Fort McKay, Alberta

International Nuclear Information System (INIS)

Faisal, K.; Gamal El-Din, M.

2006-01-01

Ambient air trends were assessed using data collected over an 8 year period from monitoring stations in Edmonton and Fort McKay, Alberta. In particular, the study evaluated the short term trends in the concentration of carbon monoxide (CO), nitrogen dioxide (NO 2 ), ozone (O 3 ), and particulate matter (PM 2.5 ) in Edmonton, as well as the NO 2 , O 3 , PM 2.5 , and total hydrocarbons in Fort McKay. In order to evaluate the ambient air trends, this study examined the changes in concentrations of these pollutants between the 50 - 90 percentiles of concentration distributions for a calendar year. These statistics were assumed to be linear over the period of study and fitted using simple linear regression. Hypothesis tests were performed to determine if the slopes of the best-fit lines were greater or less than zero. There was no indication of a statistically significant short-term trend for NO 2 and O 3 for the city of Edmonton. However, statistically pronounced decreasing trends were noted for CO and PM 2.5 . There was no indication of statistically significant trend for any of the pollutants examined at Fort McKay over the study period. It was cautioned that since the period of study over which trends were examined was short, the changes or lack of changes observed do not necessarily indicate long term trends. However, the results suggest that air quality has remained unchanged during the last 6 to 8 years, despite increased economic development in Edmonton and continued oil sands development in Fort McKay

Cultural keystone species in oil sands mine reclamation, Fort McKay, Alberta, Canada

Energy Technology Data Exchange (ETDEWEB)

Garibaldi, A.; Straker, J. [Stantec Ltd., Sidney, BC (Canada)

2009-07-01

Cultural keystone species (CKS) shape the cultural identify of people through the roles they have in diet, material and spiritual practices. The use of the CKS concept is regarded as a method of addressing linked social and ecological issues. This paper presented the results of using the CKS model in the indigenous community of Fort McKay, Alberta to address, social, ecological and spiritual values in regional mine-land reclamation. Fort McKay is at the epicenter of the existing mine developments. Its residents regard human and environmental health to be be linked and therefore experience the effects of development and subsequent reclamation on both cultural and ecological levels. The community is actively engaged in working with the local mining companies on issues of mine reclamation design. In order to hold meaning to the local people, oil sand operators used the CKS concept in their reclamation efforts to take into account ecological functionality and also address the linked social factors. Five CKS were identified through a literature review and extensive community interviews. The list includes moose, cranberry, blueberry, ratroot and beaver. These 5 CKS were used to focus discussions and make recommendations for relevant land reclamation within Fort McKay traditional territory. The project has influenced the way both the community and oil sands operators engage with reclamation. Lessons learned from this process will help direct reclamation activities on other portions of traditional territory, while offering guidance to other regional developers for addressing cultural values in reclamation on their leases. 15 refs., 1 fig.

Drilling and Testing the DOI041A Coalbed Methane Well, Fort Yukon, Alaska

Science.gov (United States)

Clark, Arthur; Barker, Charles E.; Weeks, Edwin P.

2009-01-01

The need for affordable energy sources is acute in rural communities of Alaska where costly diesel fuel must be delivered by barge or plane for power generation. Additionally, the transport, transfer, and storage of fuel pose great difficulty in these regions. Although small-scale energy development in remote Arctic locations presents unique challenges, identifying and developing economic, local sources of energy remains a high priority for state and local government. Many areas in rural Alaska contain widespread coal resources that may contain significant amounts of coalbed methane (CBM) that, when extracted, could be used for power generation. However, in many of these areas, little is known concerning the properties that control CBM occurrence and production, including coal bed geometry, coalbed gas content and saturation, reservoir permeability and pressure, and water chemistry. Therefore, drilling and testing to collect these data are required to accurately assess the viability of CBM as a potential energy source in most locations. In 2004, the U.S. Geological Survey (USGS) and Bureau of Land Management (BLM), in cooperation with the U.S. Department of Energy (DOE), the Alaska Department of Geological and Geophysical Surveys (DGGS), the University of Alaska Fairbanks (UAF), the Doyon Native Corporation, and the village of Fort Yukon, organized and funded the drilling of a well at Fort Yukon, Alaska to test coal beds for CBM developmental potential. Fort Yukon is a town of about 600 people and is composed mostly of Gwich'in Athabascan Native Americans. It is located near the center of the Yukon Flats Basin, approximately 145 mi northeast of Fairbanks.

Nuclear power for coexistence with nature, high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Kaneko, Yoshihiko

1996-01-01

Until this century, it is sufficient to aim at the winner of competition in human society to obtain resources, and to entrust waste to natural cleaning action. However, the expansion of social activities has been too fast, and the scale has become too large, consequently, in the next century, the expansion of social activities will be caught by the structure of trilemma that is subjected to the strong restraint and selection from the problems of finite energy and resources and environment preservation. In 21st century, the problems change to those between mankind and nature. Energy supply and population increase, envrionment preservation and human activities, and the matters that human wisdom should bear regarding energy technology are discussed. In Japan, the construction of the high temperature engineering test reactor (HTTR) is in progress. The design of high temperature gas-cooled reactors and their features on the safety are explained. The capability of reducing CO 2 release of high temperature gas-cooled reactors is reported. In future, it is expected that the time of introducing high temperature gas-cooled reactors will come. (K.I.)

Socioeconomic impacts of nuclear generating stations: St. lucie case study. Technical report 1 Oct 78-4 Jan 82

International Nuclear Information System (INIS)

Weisiger, M.L.; Pijawka, K.D.

1982-07-01

The report documents a case study of the socioeconomic impacts of the construction and operation of the St. Lucie nuclear power station. It is part of a major post-licensing study of the socioeconomic impacts at twelve nuclear power stations. The case study covers the period beginning with the announcement of plans to construct the reactor and ending in the period, 1980-81. The case study deals with changes in the economy, population, settlement patterns and housing, local government and public services, social structure, and public response in the study are during the construction/operation of the reactor. A regional modeling approach is used to trace the impact of construction/operation on the local economy, labor market, and housing market. Emphasis in the study is on the attribution of socioeconomic impacts to the reactor or other causal factors. As part of the study of local public response to the construction/operation of the reactor, the effects of the Three Mile Island accident are examined

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...

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…

Fast reactor development and worldwide cooperation in Generation-IV International Forum

International Nuclear Information System (INIS)

Sagayama, Yutaka

2013-01-01

Objectives of Gen-IV systems development: Goals: Four challenging technology goals have been defined to be applied to innovative nuclear reactor concepts in the 21st century: 1) Safety and Reliability (safe and reliable operation, no offsite emergency response); 2) Sustainability (effective fuel utilization, minimization of nuclear waste); 3) Proliferation Resistance & Physical Protection (to assure unattractive and the least desirable route for diversion or theft of weapons-usable materials, and provide increased physical protection against acts of terrorism); 4) Economic Competitiveness (life-cycle cost advantage over other energy resources). Phase: Each Generation-IV reactor system is one of three stages. 1) Viability Phase; 2) Performance Phase; 3) Demonstration Phase. Target: Commercial Deployment is expected around 2030s or beyond

Medline Plus

Full Text Available ... Orthopedics Center, State College, PA, 8/21/2014) Rotator Cuff Injuries Shoulder Arthroscopy (St. Francis Eastside Hospital, Greenville, ... Medical Center, Fort Lauderdale, FL, 11/04/2011) Rotator Cuff Injuries Shoulder Arthroscopy (St. Francis Eastside Hospital, Greenville, ...

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...

Transactions of the Twenty-First Water Reactor Safety Information Meeting

International Nuclear Information System (INIS)

Monteleone, S.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session

Transactions of the Twenty-First Water Reactor Safety Information Meeting

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.

1993-10-01

This report contains summaries of papers on reactor safety research to be presented at the 21st Water Reactor Safety Information Meeting at the Bethesda Marriott Hotel, Bethesda, Maryland, October 25--27, 1993. The summaries briefly describe the programs and results of nuclear safety research sponsored by the Office of Nuclear Regulatory Research, US NRC. Summaries of invited papers concerning nuclear safety issues from US government laboratories, the electric utilities, the Electric Power Research Institute (EPRI), the nuclear industry, and from foreign governments and industry are also included. The summaries have been compiled in one report to provide a basis for meaningful discussion and information exchange during the course of the meeting and are given in the order of their presentation in each session.

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...

Efficacy and Safety Evaluation of Myostaal Forte, a Polyherbal Formulation, in Treatment of Knee Osteoarthritis: A Randomised Controlled Pilot Study

Directory of Open Access Journals (Sweden)

Raakhi K Tripathi

2017-10-01

Full Text Available Introduction: Myostaal Forte, a proprietary poly-herbal formulation, is mixture of nine herbal plant extracts which possess analgesic, anti-inflammatory and chondroprotective properties. Aim: A prospective, randomised, active controlled, 2-arm, parallel group, assessor blind study was planned to evaluate clinical efficacy and safety of Myostaal Forte in patients of knee osteoarthritis. Materials and Methods: Idiopathic knee osteoarthritis cases as per American College of Rheumatology (ACR clinical criteria were screened and recruited. A total of sixty patients were assigned to receive Myostaal Forte TDS (n=30 or Paracetamol 650 mg TDS (n=30 for six weeks. Naproxen was rescue analgesia. Modified Western Ontario and McMaster Universities Arthritis Index (WOMAC, Visual Analogue Scale (VAS, global assessment scores determined by orthopaedic physician at baseline, two, four, six weeks and telephonically at eight weeks. Safety was assessed through laboratory investigations at baseline and six weeks, adverse events and tolerability. Data were expressed as Mean±SD and analysed by Chi-square and unpaired t-test. p0.05. No significant adverse events, changes in the laboratory parameters and excellent compliance to treatment were seen in both the groups. Conclusion: Earlier onset analgesic effect with sustained chondroprotection after treatment cessation makes Myostaal Forte, a safe and effective alternative for treatment of knee osteoarthritis.

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.

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.

Biomechanical testing of zirconium dioxide osteosynthesis system for Le Fort I advancement osteotomy fixation.

Science.gov (United States)

Hingsammer, Lukas; Grillenberger, Markus; Schagerl, Martin; Malek, Michael; Hunger, Stefan

2018-01-01

The following work is the first evaluating the applicability of 3D printed zirconium dioxide ceramic miniplates and screws to stabilize maxillary segments following a Le-Fort I advancement surgery. Conventionally used titanium and individual fabricated zirconium dioxide miniplates were biomechanically tested and compared under an occlusal load of 120N and 500N using 3D finite element analysis. The overall model consisted of 295,477 elements. Under an occlusal load of 500N a safety factor before plastic deformation respectively crack of 2.13 for zirconium dioxide and 4.51 for titanium miniplates has been calculated. From a biomechanical point of view 3D printed ZrO 2 mini-plates and screws are suggested to constitute an appropriate patient specific and metal-free solution for maxillary stabilization after Le Fort I osteotomy. Copyright © 2017 Elsevier Ltd. All rights reserved.

A non-conventional procedure for the 3D modeling of WWI forts

Directory of Open Access Journals (Sweden)

E. Nocerino

2014-06-01

Full Text Available 2014 is the hundredth anniversary of the outbreak of the First World War (WWI – or Great War – in Europe and a number of initiatives have been planned to commemorate the tragic event. Until 1918, the Italian Trentino – Alto Adige region was under the Austro – Hungarian Empire and represented one of the most crucial and bloody war front between the Austrian and Italian territories. The region borders were constellated of military fortresses, theatre of battles between the two opposite troops. Unfortunately, most of these military buildings are now ruined and their architectures can be hardly appreciated. The paper presents the initial results of the VAST project (VAlorizzazione Storia e Territorio – Valorization of History and Landscape, that aims to digitally reconstruct the forts located on the plateaus of Luserna, Lavarone and Folgaria. An integrated methodology has been adopted to collect and employ all possible source of information in order to derive precise and photo-realistic 3D digital representations of WWI forts.

Prospects for competitive nuclear power into the 21st century

International Nuclear Information System (INIS)

Shapar, H.K.; Sasaki, T.; Thexton, H.E.

1986-10-01

Nuclear power stations committed today will be commissioned in the mid- to late-1990s and will operate for most of their lives in the 21st century. Utilities considering the nuclear option for new increments of capacity are, therefore, required to make judgements now on the competitiveness of nuclear plants well beyond the turn of the century. Reactors committed for mid-1990 startup could have a lifetime competitive cost advantage of 20% to 80% over coal-fired plants in most NEA areas studied except for some parts of North America where low cost coal is available. Nuclear plants would retain an economic advantage in most countries even if they were to be used for load-following with resultant lifetime capacity (load) factors as low as about 60% (or even lower in many countries). Uranium resources are sufficiently large that fuel supply should not constrain nuclear power development well into the 21st century, as long as uranium prices provide a market incentive for continued exploration and mine development. Uranium prices seem unlikely to rise to a level which would remove nuclear's advantage within the next several decades. There is apparently no technological reason why the ''back-end'' of the fuel cycle (spent fuel transportation, storage, reprocessing and disposal of high level waste) and decommissioning of reactors should constrain further deployment of nuclear power. While the costs of these activities appear high in absolute terms, they will be relatively low per unit of electricity generated. Thus, even though there remain uncertainties regarding these future costs, they should have little impact on electricity consumers. (author)

A Study to Evaluate the Organization and the Operating Procedures of the Patient Assistance Function at Brooke Army Medical Center, Fort Sam Houston, Texas

Science.gov (United States)

1979-08-01

15 March 1979. 59Interview with Wendy L. Farace , Head Nurse, Obstetrics/Gynecology Clinic, Brooke Army Medical Center, Fort Sam Houston, Texas, 8...6 February 1979. Farace , Wendy L. Head Nurse, Obstetrica/Gynecology Clinic, Brooke Army Medical Center, Fort Sam Houston, Texas. Interview, 8 January

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...

Postkort fra fortiden afslører romerske soldaters hverdag

DEFF Research Database (Denmark)

Evers, Kasper Grønlund

2014-01-01

Selvom det romerske fort Vindolanda lige syd for Hadrians mur i Nordengland var kendt for sine velbevarede ruiner, kunne ingen før 1973 have forudset, at stedet skulle gemme på dét, der siden er blevet udnævnt til Storbritanniens største historiske skat.......Selvom det romerske fort Vindolanda lige syd for Hadrians mur i Nordengland var kendt for sine velbevarede ruiner, kunne ingen før 1973 have forudset, at stedet skulle gemme på dét, der siden er blevet udnævnt til Storbritanniens største historiske skat....

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...

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...

Warm Dry Weather Conditions Cause of 2016 Fort McMurray Wild Forest Fire and Associated Air Quality

Science.gov (United States)

de Azevedo, S. C.; Singh, R. P.; da Silva, E. A., Sr.

2016-12-01

The climate change is evident from the increasing temperature around the world, day to day life and increasing frequency of natural hazards. The warm and dry conditions are the cause of frequent forest fires around the globe. Forest fires severely affect the air quality and human health. Multi sensor satellites and dense network of ground stations provide information about vegetation health, meteorological, air quality and atmospheric parameters. We have carried out detailed analysis of satellite and ground data of wild forest fire that occurred in May 2016 in Fort McMurray, Alberta, Canada. This wild forest fire destroyed 10 per cent of Fort McMurray's housing and forced more than 90,000 people to evacuate the surrounding areas. Our results show that the warm and dry conditions with low rainfall were the cause of Fort McMurray wild fire. The air quality parameters (particulate matter, CO, ozone, NO2, methane) and greenhouse gases measured from Atmospheric Infrared Sounder (AIRS) satellite show enhanced levels soon after the forest fire. The emissions from the forest fire affected health of population living in surrounding areas up to 300 km radius.

Implementation of digital control and protection systems of China advanced research reactor

International Nuclear Information System (INIS)

Zeng Hai; Jin Huajin; Xu Qiguo; Zhang Mingkui

2005-01-01

China Advanced Research Reactor (CARR), a reactor of the 21st century with high performance is being constructed in China. The requirements of reliability and stability on the control and protection (c and p) system are the main points raised. Especially, with the development of digital technology, the c and p system of CARR is demanded to match the trend of digitization in the field of reactor control. The c and p system, including reactor protection system, reactor monitoring and control system, reactor power regulating system, and the mitigation system for ATWS (Anticipate Transient Without Scram), adopts digital technology, and the digital display screen will replace the analog panels in the main control room. The c and p system of CARR adopts redundant technology with 2 or 3 redundant channels to improve the system reliability. The 10/100 Mbps self-adaptive redundant optic fiber industry Ethernet ring network is used to interlink operator workstations, supervisor workstation, and I/O control stations. Commercial grade equipment with mature experience in industrial application are applied to the c and p system of CARR, which have high reliability, good interchangeability, and is easily purchased, the software-developing tools fully match the international industry standards. The realization of digital c and p system of CARR will promote the progress of digital control technology for reactors in China, and certainly become a technical basic platform for developing informational and intelligent reactors in China. (authors)

Stāstu stāstīšanas izmantošana lasīšanas veicināšanas pasākumos

OpenAIRE

Mežjāne, Signe

2014-01-01

Bakalaura darba mērķis ir noskaidrot, vai stāstu stāstīšana ir efektīvs lasīšanas veicināšanas paņēmiens bērniem. Pētījuma problēma ir mūsdienu bērnu nevēlēšanās lasīt, lasīšanu varētu veicināt stāstu stāstīšana bibliotēkās. Darbā tiek analizēts stāstu stāstīšanas process atbilstoši bērnu vecumposmam, izmantojot dažādus runas stilus. Teorētiskā bāze ir balstīta uz Lasītāja reakcijas kritikas teoriju un Runas darbības teoriju. Galvenie pētījuma uzdevumi ir noskaidrot, vai Latvijas bibliotē...

Burnup dependent core neutronic calculations for research and training reactors via SCALE4.4

International Nuclear Information System (INIS)

Tombakoglu, M.; Cecen, Y.

2001-01-01

In this work, the full core modelling is performed to improve neutronic analyses capability for nuclear research reactors using SCALE4.4 code system. KENOV.a module of SCALE4.4 code system is utilized for full core neutronic analysis. The ORIGEN-S module is coupled with the KENOV.a module to perform burnup dependent neutronic analyses. Results of neutronic calculations for 1 st cycle of Cekmece TR-2 research reactor are presented. In particular, coupling of KENOV.a and ORIGEN-S modules of SCALE4.4 is discussed. The preliminary results of 2-D burnup dependent neutronic calculations are also given. These results are extended to burnup dependent core calculations of TRIGA Mark-II research reactors. The code system developed here is similar to the code system that couples MCNP and ORIGEN2.(author)

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

[Russian experience with Vitaprost Forte suppositories in patients with lower urinary tract symptoms and benign prostatic hyperplasia: comparative analysis of studies].

Science.gov (United States)

Korneev, I A

2017-07-01

The article reviews the domestic studies showing the efficacy and safety of suppositories containing prostate extract (Samprost substance) Vitaprost Forte in treating men with lower urinary tract symptoms secondary to benign prostatic hyperplasia. The data obtained by Russian specialists confirm the effectiveness of Vitaprost Forte suppositories in managing patients with moderate LUTS and infravesical obstruction caused by BPH to reduce dysuria, improve the quality of life and normalize urodynamic parameters.

75 FR 51945 - Safety Zone; Potomac River, St. Mary's River, St. Inigoes, MD

Science.gov (United States)

2010-08-24

...-AA00 Safety Zone; Potomac River, St. Mary's River, St. Inigoes, MD AGENCY: Coast Guard, DHS. ACTION... of the St. Mary's River, a tributary of the Potomac River. This action is necessary to provide for.... Navy helicopter located near St. Inigoes, Maryland. This safety zone is intended to protect the...

Sustainable Materials Replacement for Prevention of Corrosion at Fort Lewis, WA

Science.gov (United States)

2009-08-01

documents a building reclamation project at Fort Lewis, WA, in which significant portions of the work were completed using market -available sustainable...contractors to submit previous technology performance or test documentation, which is generally published as products are introduced to the market . If...evaluate the performance of the E2C2 in its final configuration. Energy use will be analyzed using a Building Infor- mation Model ( BIM ) to simulate

78 FR 33808 - Foreign-Trade Zone (FTZ) 39-Dallas-Fort Worth, Texas; Notification of Proposed Production...

Science.gov (United States)

2013-06-05

... Street, and 4600 Blue Mound Road, Fort Worth (Tarrant County), Texas. A separate application for ``usage... abroad include: plastic labels; parts of fans (housings, grills, pedestal assemblies, blades); electric...

Ânion gap corrigido para albumina, fosfato e lactato é um bom preditor de íon gap forte em pacientes enfermos graves: estudo de coorte em nicho

Directory of Open Access Journals (Sweden)

Fernando Godinho Zampieri

2013-09-01

Full Text Available OBJETIVO: Ânion gap corrigido e íon gap forte são usados comumente para estimar os ânions não medidos. Avaliamos o desempenho do ânion gap corrigido para albumina, fosfato e lactato na predição do íon gap forte em uma população mista de pacientes enfermos graves. Formulamos a hipótese de que o ânion gap corrigido para albumina, fosfato e lactato seria um bom preditor do íon gap forte, independentemente da presença de acidose metabólica. Além disso, avaliamos o impacto do íon gap forte por ocasião da admissão na mortalidade hospitalar. MÉTODOS: Incluímos 84 pacientes gravemente enfermos. A correlação e a concordância entre o ânion gap corrigido para albumina, fosfato e lactato e o íon gap forte foi avaliada utilizando-se os testes de correlação de Pearson, regressão linear, plot de Bland-Altman e pelo cálculo do coeficiente de correlação interclasse. Foram realizadas duas análises de subgrupos: uma para pacientes com excesso de base -2mEq/L (grupo com alto excesso de base. Foi realizada uma regressão logística para avaliar a associação entre os níveis de íon gap forte na admissão e a mortalidade hospitalar. RESULTADOS: Houve correlação muito forte e uma boa concordância entre o ânion gap corrigido para albumina, fosfato e lactato e o íon gap forte na população geral (r2=0,94; bias 1,40; limites de concordância de -0,75 a 3,57. A correlação foi também elevada nos grupos com baixo excesso de base (r2=0,94 e alto excesso de base (r2=0,92. Estavam presentes níveis elevados de íon gap forte em 66% da população total e 42% dos casos do grupo alto excesso de. Íon gap forte não se associou com a mortalidade hospitalar, conforme avaliação pela regressão logística. CONCLUSÃO: O ânion gap corrigido para albumina, fosfato e lactato e o íon gap forte tiveram uma excelente correlação. Os ânions não medidos estão frequentemente elevados em pacientes gravemente enfermos com excesso de base

Vibration monitoring/diagnostic techniques, as applied to reactor coolant pumps

International Nuclear Information System (INIS)

Sculthorpe, B.R.; Johnson, K.M.

1986-01-01

With the increased awareness of reactor coolant pump (RCP) cracked shafts, brought about by the catastrophic shaft failure at Crystal River number3, Florida Power and Light Company, in conjunction with Bently Nevada Corporation, undertook a test program at St. Lucie Nuclear Unit number2, to confirm the integrity of all four RCP pump shafts. Reactor coolant pumps play a major roll in the operation of nuclear-powered generation facilities. The time required to disassemble and physically inspect a single RCP shaft would be lengthy, monetarily costly to the utility and its customers, and cause possible unnecessary man-rem exposure to plant personnel. When properly applied, vibration instrumentation can increase unit availability/reliability, as well as provide enhanced diagnostic capability. This paper reviews monitoring benefits and diagnostic techniques applicable to RCPs/motor drives

Freight Advanced Traveler Information System (FRATIS) - Dallas-Fort Worth : as-built system architecture and design.

Science.gov (United States)

This document describes the As-Built System Architecture and Design 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 specif...

When a plant shuts down: The psychology of decommissioning

International Nuclear Information System (INIS)

Schulz, J.; Crawford, A.C.

1993-01-01

Within the next decade, 10 to 25 nuclear plants in the United States may be taken off line. Many will have reached the end of their 40-year life cycles, but others will be retired because the cost of operating them has begun to outweigh their economic benefit. Such was the case at Fort St. Vrain, the first decommissioning of a US commercial plant under new Nuclear Regulatory Commission (NRC) regulations. Two major problems associated with decommissioning plants have been obvious: (1) the technical challenges and costs of decommissioning, and (2) the cost of maintaining and finally decommissioning a plant after a safe storage (SAFSTOR) period of approximately 60 years. What has received little attention is the challenge that affects not only the people who make a plant work, but the quality of the solutions to these problems: how to maintain effective organizational performance during the process of downsizing and decommissioning and/or SAFSTOR. The quality of technical solutions for closing a plant, as well as how successfully the decommissioning process is held within or below budget, will depend largely on how effectively the nuclear organization functions as a social unit. Technical and people issues are bound together. The difficulty is how to operate a plant effectively when plant personnel have no sense of long-term security. As the nuclear power industry matures and the pace for closing operating plants accelerates, the time has come to prepare for the widespread decommissioning of plants. The industry would be well served by conducting a selective, industry-wide evaluation of plants to assess its overall readiness for the decommissioning process. A decommissioning is not likely to be trouble free, but with a healthy appreciation for the human side of the process, it will undoubtedly go more smoothly than if approached as a matter of dismantling a machine

Technical evolution and operation of French CO2 cooled reactors (UNGG)

International Nuclear Information System (INIS)

Berthion, Y.

1986-10-01

The technical evolution of the five French CO 2 cooled reactors (UNGG) from 1981 to 1986 needs to be outlined. These technical evolutions concerned the fuel element of Bugey 1 which is now slightly enriched, as well as the load reduction operation required by the grid. In addition work in underway to increase the safety at the two St Laurent units, or to repair the hot steel upper-structures of Chinon-3 unit

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.

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.

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.

Energy Optimization Assessment at U.S. Army Installations: Fort Bliss, TX

Science.gov (United States)

2008-09-01

Log dampers, temperatures, actuator signals, and other parameters to identify problems. Adjust chiller and boiler setpoints and control curves...installation. The lowest setpoints were found in the Centennial Club, with 52 °F during unoccupied hours (morn- ing). The chillers ran pretty much fully loaded...ER D C/ CE R L TR -0 8 -1 5 Energy Optimization Assessment at U.S. Army Installations Fort Bliss, TX David M. Underwood, Alexander M

Data Mining the Corporate Dental System of USA DENTAC Fort Bragg

Science.gov (United States)

2016-06-10

appointment data were queried for active duty Soldiers assigned to Fort Bragg, NC. All data were analyzed by using SPSS version 22.0 ( SPSS , Chicago, IL...Clinic. For each appointment the appointment type, date, and dental wellness classification were retained for analysis . For the purposes of this study...considered part of the Go First Class program. The analysis was split into multiple phases, beginning with a chi-square test for trend to

Current status and prospects of research reactors

International Nuclear Information System (INIS)

Gabaraev, A.B.; Cherepnin, Yu.S.; Tretyakov, I.T.; Khmelshikov, V.V.; Dollezhal, N.A.

2009-01-01

Full text: The first nuclear research reactors (RR) appeared in the 1940s. Their initial purpose was to provide knowledge of the main processes associated with neutron-induced nuclear reactions. Later, the rang of problems addressed expanded substantially. Besides fundamental research in the properties of matter, such reactors are successfully used for dealing with problems in the fields of materials science, nuclear engineering, medicine, isotope production, education, etc. Over the whole period of RR fleet growth, more than six hundred nuclear research facilities were built in 70 countries of the world. As of the end of 2008, the number of Russian research reactors in service was about 20% of the globally operating RR fleet. This paper discusses the current status of the world's RR fleet and describes the capabilities of the experimental reactor facilities existing in Russia. In the 21st century, research reactors will remain in demand to solve scientific and technological problems for innovative development of society. The emerging renaissance of nuclear power, the expanding RR uses for production of isotopes and other applications, the increase in the number of countries willing to use nuclear technologies in energy production, industry and science - all contribute to a rebirth of interest in research reactors. One of the ways to improve the experimental capabilities lies in radical upgrading of the reactor facilities with qualitative changes in the main neutronic characteristics of the core. The associated design approaches are illustrated with the example of the IBR-2M reactor at the JNRI in Dubna. The imperative need restricting the spread of nuclear threat leads us to give up using highly enriched uranium in most research reactors. Development of RR fuel with reduced enrichment in uranium has been one of the priority objectives of NIKIET for many years. This paper presents the latest results obtained along these lines, as applied to pool-type research

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.

Vegetation inventory, mapping, and classification report, Fort Bowie National Historic Site

Science.gov (United States)

Studd, Sarah; Fallon, Elizabeth; Crumbacher, Laura; Drake, Sam; Villarreal, Miguel

2013-01-01

A vegetation mapping and characterization effort was conducted at Fort Bowie National Historic Site in 2008-10 by the Sonoran Desert Network office in collaboration with researchers from the Office of Arid lands studies, Remote Sensing Center at the University of Arizona. This vegetation mapping effort was completed under the National Park Service Vegetation Inventory program which aims to complete baseline mapping inventories at over 270 national park units. The vegetation map data was collected to provide park managers with a digital map product that met national standards of spatial and thematic accuracy, while also placing the vegetation into a regional and even national context. Work comprised of three major field phases 1) concurrent field-based classification data collection and mapping (map unit delineation), 2) development of vegetation community types at the National Vegetation Classification alliance or association level and 3) map accuracy assessment. Phase 1 was completed in late 2008 and early 2009. Community type descriptions were drafted to meet the then-current hierarchy (version 1) of the National Vegetation Classification System (NVCS) and these were applied to each of the mapped areas. This classification was developed from both plot level data and censused polygon data (map units) as this project was conducted as a concurrent mapping and classification effort. The third stage of accuracy assessment completed in the fall of 2010 consisted of a complete census of each map unit and was conducted almost entirely by park staff. Following accuracy assessment the map was amended where needed and final products were developed including this report, a digital map and full vegetation descriptions. Fort Bowie National Historic Site covers only 1000 acres yet has a relatively complex landscape, topography and geology. A total of 16 distinct communities were described and mapped at Fort Bowie NHS. These ranged from lush riparian woodlands lining the

Procarti Forte in the Complex Treatment of Patients with Early-Stage Osteoarthritis

Directory of Open Access Journals (Sweden)

O.A. Burianov

2016-04-01

Full Text Available The article deals with the issue of the treatment of osteoarthritis. The review of current recommendations on the feasibility of using glucosamine sulfate, chondroitin sulfate, hyaluronic acid, using of SYSADOA drugs, metabolic drugs was performed. The study on the efficacy and safety of using combination drug Procarti Forte in the system of treatment of patients with early-stage osteoarthritis is presented.

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...

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...

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...

Proposal for Dual Pressurized Light Water Reactor Unit Producing 2000 MWe

International Nuclear Information System (INIS)

Kang, Kyoung Min; Noh, Sang Woo; Suh, Kune Yull

2009-01-01

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the 21 st century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well

International conference on opportunities and challenges for water cooled reactors in the 21. century. PowerPoint presentations

International Nuclear Information System (INIS)

2009-01-01

Water Cooled Reactors have been the keystone of the nuclear industry in the 20th Century. As we move into the 21st Century and face new challenges such as the threat of climate change or the large growth in world energy demand, nuclear energy has been singled out as one of the sources that could substantially and sustainably contribute to power the world. As the nuclear community worldwide looks into the future with the development of advanced and innovative reactor designs and fuel cycles, it becomes important to explore the role Water Cooled Reactors (WCRs) will play in this future. To support the future role of WCRs, substantial design and development programmes are underway in a number of Member States to incorporate additional technology improvements into advanced nuclear power plants (NPPs) designs. One of the key features of advanced nuclear reactor designs is their improved safety due to a reduction in the probability and consequences of accidents and to an increase in the operator time allowed to better assess and properly react to abnormal events. A systematic approach and the experience of many years of successful operation have allowed designers to focus their design efforts and develop safer, more efficient and more reliable designs, and to optimize plant availability and cost through improved maintenance programs and simpler operation and inspection practices. Because many of these advanced WCR designs will be built in countries with no previous nuclear experience, it is also important to establish a forum to facilitate the exchange of information on the infrastructure and technical issues associated with the sustainable deployment of advanced nuclear reactors and its application for the optimization of maintenance of operating nuclear power plants. This international conference seeks to be all-inclusive, bringing together the policy, economic and technical decision-makers and the stakeholders in the nuclear industry such as operators, suppliers

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

The Coast Artillery Journal. Volume 79, Number 4, July-August 1936

Science.gov (United States)

1936-08-01

to play, and to re- gard manhood as something to be acquired in the nebulous future. Comes a new face to the ROTC; its owner grins a friendly grin...Department Artillery Officer COLONEL LEWIS TURTLE , CA.C. Fort Amador COLONEL EARLE D’A. PEARCE 4thCA. (AA) F011 Sherman COLONEL WILLIAM M. CoLVIN 1stCA. Fort

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

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,

Dust Plume Modeling from Ranges and Maneuver Areas on Fort Bliss and the White Sands Missile Range: Final Report

Energy Technology Data Exchange (ETDEWEB)

Chapman, Elaine G.; Barnard, James C.; Rutz, Frederick C.; Pekour, Mikhail S.; Rishel, Jeremy P.; Shaw, William J.

2009-05-04

The potential for air quality impacts from heavy mechanized vehicles operating on and between the unpaved main supply routes at Fort Bliss and White Sands Missile Range was investigated. This report details efforts by the staff of Pacific Northwest National Laboratory for the Fort Bliss Directorate of Environment in this investigation. Dust emission and dispersion from typical move-out activities occurring on the installations were simulated using the atmospheric modeling system DUSTRAN. Major assumptions associated with designing the modeling scenarios are summarized and results of simulations conducted under these assumptions are presented for four representative meteorological periods.

Assessment of soil-gas, seep, and soil contamination at the North Range Road Landfill, Fort Gordon, Georgia, 2008-2009

Science.gov (United States)

Landmeyer, James E.; Falls, W. Fred; Ratliff, W. Hagan; Wellborn, John B.

2011-01-01

Soil gas, seeps, and soil were assessed for contaminants at the North Range Road Landfill at Fort Gordon, Georgia, from October 2008 to September 2009. The assessment included delineating organic contaminants present in soil-gas samples beneath the area estimated to be the landfill and in water samples collected from three seeps at the base of the landfill. Inorganic contaminants were determined in three seep samples and in soil samples. This assessment was conducted to provide environmental contamination data to Fort Gordon pursuant to requirements for the Resource Conservation and Recovery Act Part B Hazardous Waste Permit process.

Status and some safety philosophies of the China advanced research reactor CARR

International Nuclear Information System (INIS)

Luzheng Yuan

2001-01-01

The existing two research reactors, HWRR (heavy water research reactor) and SPR (swimming pool reactor), have been operated by China Institute of Atomic Energy (CIAE) since, respectively, 1958 and 1964, and are both in extending service and facing the aging problem. It is expected that they will be out of service successively in the beginning decade of the 21 st century. A new, high performance and multipurpose research reactor called China advanced research reactor (CARR) will replace these two reactors. This new reactor adopts the concept of inverse neutron trap compact core structure with light water as coolant and heavy water as the outer reflector. Its design goal is as follows: under the nuclear power of 60MW, the maximum unperturbed thermal neutron flux in peripheral D 2 O reflector not less than 8 x 10 14 n/cm 2 . s while in central experimental channel, if the central cell to be replaced by an experimental channel, the corresponding value not less than 1 x 10 15 n/cm 2 . s. The main applications for this research reactor will cover RI production, neutron scattering experiments, NAA and its applications, neutron photography, NTD for monocrystaline silicon and applications on reactor engineering technology. By the end of 1999, the preliminary design of CARR was completed, then the draft of preliminary safety analysis report (PSAR) was submitted to the relevant authority at the end of 2000 for being reviewed. Now, the CARR project has entered the detail design phase and safety reviewing procedure for obtaining the construction permit from the relevant licensing authority. This paper will only briefly introduce some aspects of safety philosophy of CARR design and PSAR. (orig.)

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

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...

76 FR 55702 - National Register of Historic Places; Notification of Pending Nominations and Related Actions

Science.gov (United States)

2011-09-08

..., MPS) Roughly bounded by Terry St., Ward Ave., Ellis St. & Darby Ave., Fort Smith, 11000693 Washington... Washington County Smith, E.L., Roundhouse Granite Shed, 23 Burnham Meadows, Barre, 11000704 Request for REMOVAL has been made for the following resources: ARKANSAS Howard County Boyd, Adam, House E. of Center...

End-use energy characterization and conservation potentials at DoD Facilities: An analysis of electricity use at Fort Hood, Texas

Energy Technology Data Exchange (ETDEWEB)

Akbari, H.; Konopacki, S.

1995-05-01

This report discusses the application of the LBL`s End-use Disaggregation Algorithm (EDA) to a DoD installation and presents hourly reconciled end-use data for all major building types and end uses. The project initially focused on achieving these objectives and pilot-testing the methodology at Fort Hood, Texas. Fort Hood, with over 5000 buildings was determined to have representative samples of nearly all of the major building types in use on DoD installations. These building types at Fort Hood include: office, administration, vehicle maintenance, shop, hospital, grocery store, retail store, car wash, church, restaurant, single-family detached housing, two and four-plex housings, and apartment building. Up to 11 end uses were developed for each prototype, consisting of 9 electric and 2 gas; however, only electric end uses were reconciled against known data and weather conditions. The electric end uses are space cooling, ventilation, cooking, miscellaneous/plugs, refrigeration, exterior lighting, interior lighting, process loads, and street lighting. The gas end uses are space heating and hot water heating. Space heating energy-use intensities were simulated only. The EDA was applied to 10 separate feeders from the three substations at Fort Hood. The results from the analyses of these ten feeders were extrapolated to estimate energy use by end use for the entire installation. The results show that administration, residential, and the bar-rack buildings are the largest consumers of electricity for a total of 250GWh per year (74% of annual consumption). By end use, cooling, ventilation, miscellaneous, and indoor lighting consume almost 84% of total electricity use. The contribution to the peak power demand is highest by residential sector (35%, 24 MW), followed by administration buildings (30%), and barrack (14%). For the entire Fort Hood installation, cooling is 54% of the peak demand (38 MW), followed by interior lighting at 18%, and miscellaneous end uses by 12%.

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...

Summary of the 3rd workshop on the reduced-moderation water reactor

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

Summary of the 3rd workshop on the reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

Study of a compact reversed shear Tokamak reactor

International Nuclear Information System (INIS)

Okano, K.; Asaoka, Y.; Tomabechi, K.; Yoshida, T.; Hiwatari, R.; Ogawa, Y.; Tokimatsu, K.; Yamamoto, T.; Inoue, N.; Murakami, Y.

1998-01-01

A reversed shear configuration, which was observed recently in some tokamak experiments, might have a possibility to realize compact and cost-competitive tokamak reactors. In this study, a compact (low cost) commercial reactor based on the shear reversed high beta equilibrium with β N =5.5, is considered, namely the compact reversed shear tokamak, CREST-1. The CREST-1 is designed with a moderate aspect ratio (R/a=3.4), which will allow us to experimentally develop this CREST concept by ITER. This will be very advantageous with regard to the fusion development strategy. The current profile for the reversed shear operation is sustained and controlled in steady state by bootstrap (88%), beam and r driven currents, which are calculated by a neo-classical model code in 3D geometry. The MHD stability has been checked by an ideal MHD stability analysis code (ERATO) and it has been confirmed that the ideal low n kink, ballooning and Mercier modes are stable while a closed conductive shell is required for stability. Such a compact tokamak can be cost-competitive as an electric power source in the 21st century and it is one possible scenario in realizing a commercial fusion reactor beyond the ITER project. (orig.)

The spirit of St. Lucie: nuclear plant built on schedule

International Nuclear Information System (INIS)

Derrickson, W.B.

1984-01-01

Florida Power and Light Company currently has four nuclear units in operation with St. Lucie Unit 2 being the last to receive an operating license in June, 1983. It's sister Unit 1 received its license in 1976 and has, through 1982, compiled one of the best operating records in the United States. The full power license for St. Lucie Unit 2 was received from the Nuclear Regulatory Commission (NRC) on June 10, 1983, just six years after construction began. The industry average for construction of nuclear plants in this time period is about 10 years. The success of the St. Lucie Unit 2 project can be at least in part attributed to planning the work, accurate and timely reporting of results via valid indicators, well trained and skilled personnel, and most of all, teamwork. During the course of the project the plant was constantly on or near schedule and always ahead of industry averages. This was done despite issuance of numerous regulations by the NRC (TMI), a 1979 hurricane which did considerable damage to the Reactor Auxiliary Building, labor problems, and an NRC schedule review team that determined the best that could be done was to complete the plant a year later. The final price tag is about $1.42 billion, including ''allowance for funds used during construction''. In operation to date the post core loading test program has been completed in less than two months, enabling the plant to be put into commercial operation only two months after its original scheduled date of May 28, 1983exclamation

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

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

78 FR 19632 - Special Local Regulations; St. Thomas Carnival Watersport Activities, Charlotte Amalie Harbor; St...

Science.gov (United States)

2013-04-02

...-AA08 Special Local Regulations; St. Thomas Carnival Watersport Activities, Charlotte Amalie Harbor; St... proposes to establish a special local regulation on the waters of Charlotte Amalie Harbor in St Thomas, USVI during the St. Thomas Carnival Watersport Activities, a high speed boat race. The event is...

Rise-to-power test in High Temperature Engineering Test Reactor. Test progress and summary of test results up to 30 MW of reactor thermal power

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi

2002-08-01

The High Temperature Engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30 MW and the reactor outlet coolant temperature of 850degC/950degC. Rise-to-power test in the HTTR was performed from April 23rd to June 6th in 2000 as phase 1 test up to 10 MW in the rated operation mode, from January 29th to March 1st in 2001 as phase 2 test up to 20 MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20 MW in the high temperature test the mechanism of the reactor outlet coolant temperature becomes 850degC at 30 MW in the rated operation mode and 950degC in the high temperature test operation mode. Phase 4 rise-to-power test to achieve the thermal reactor power of 30 MW started on October 23rd in 2001. On December 7th in 2001 it was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30 MW and 850degC respectively in the single loaded operation mode in which only the primary pressurized water cooler is operating. Phase 4 test was performed until March 6th in 2002. JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests by MEXT were passed successfully with the reactor transient test at an abnormal event as a final pre-operation test. From the test results of the rise-up-power test up to 30 MW in the rated operation mode, performance of the reactor and cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely. Some problems to be solved were found through the tests. By solving them, the reactor operation with the reactor outlet coolant temperature of 950degC will be achievable. (author)

Galleria mellonella infection model demonstrates high lethality of ST69 and ST127 uropathogenic E. coli.

Directory of Open Access Journals (Sweden)

Majed F Alghoribi

Full Text Available Galleria mellonella larvae are an alternative in vivo model for investigating bacterial pathogenicity. Here, we examined the pathogenicity of 71 isolates from five leading uropathogenic E. coli (UPEC lineages using G. mellonella larvae. Larvae were challenged with a range of inoculum doses to determine the 50% lethal dose (LD50 and for analysis of survival outcome using Kaplan-Meier plots. Virulence was correlated with carriage of a panel of 29 virulence factors (VF. Larvae inoculated with ST69 and ST127 isolates (10(4 colony-forming units/larvae showed significantly higher mortality rates than those infected with ST73, ST95 and ST131 isolates, killing 50% of the larvae within 24 hours. Interestingly, ST131 isolates were the least virulent. We observed that ST127 isolates are significantly associated with a higher VF-score than isolates of all other STs tested (P≤0.0001, including ST69 (P<0.02, but one ST127 isolate (strain EC18 was avirulent. Comparative genomic analyses with virulent ST127 strains revealed an IS1 mediated deletion in the O-antigen cluster in strain EC18, which is likely to explain the lack of virulence in the larvae infection model. Virulence in the larvae was not correlated with serotype or phylogenetic group. This study illustrates that G. mellonella are an excellent tool for investigation of the virulence of UPEC strains. The findings also support our suggestion that the incidence of ST127 strains should be monitored, as these isolates have not yet been widely reported, but they clearly have a pathogenic potential greater than that of more widely recognised clones, including ST73, ST95 or ST131.

Epidemic potential of Escherichia coli ST131 and Klebsiella pneumoniae ST258: a systematic review and meta-analysis

Science.gov (United States)

Dautzenberg, M J D; Haverkate, M R; Bonten, M J M; Bootsma, M C J

2016-01-01

Objectives Observational studies have suggested that Escherichia coli sequence type (ST) 131 and Klebsiella pneumoniae ST258 have hyperendemic properties. This would be obvious from continuously high incidence and/or prevalence of carriage or infection with these bacteria in specific patient populations. Hyperendemicity could result from increased transmissibility, longer duration of infectiousness, and/or higher pathogenic potential as compared with other lineages of the same species. The aim of our research is to quantitatively estimate these critical parameters for E. coli ST131 and K. pneumoniae ST258, in order to investigate whether E. coli ST131 and K. pneumoniae ST258 are truly hyperendemic clones. Primary outcome measures A systematic literature search was performed to assess the evidence of transmissibility, duration of infectiousness, and pathogenicity for E. coli ST131 and K. pneumoniae ST258. Meta-regression was performed to quantify these characteristics. Results The systematic literature search yielded 639 articles, of which 19 data sources provided information on transmissibility (E. coli ST131 n=9; K. pneumoniae ST258 n=10)), 2 on duration of infectiousness (E. coli ST131 n=2), and 324 on pathogenicity (E. coli ST131 n=285; K. pneumoniae ST258 n=39). Available data on duration of carriage and on transmissibility were insufficient for quantitative assessment. In multivariable meta-regression E. coli isolates causing infection were associated with ST131, compared to isolates only causing colonisation, suggesting that E. coli ST131 can be considered more pathogenic than non-ST131 isolates. Date of isolation, location and resistance mechanism also influenced the prevalence of ST131. E. coli ST131 was 3.2 (95% CI 2.0 to 5.0) times more pathogenic than non-ST131. For K. pneumoniae ST258 there were not enough data for meta-regression assessing the influence of colonisation versus infection on ST258 prevalence. Conclusions With the currently available data

INPRO economic assessment of the IRIS nuclear reactor for deployment in Brazil