WorldWideScience

Sample records for experimental facilities reactor

  1. Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges

    Directory of Open Access Journals (Sweden)

    Sabharwall Piyush

    2015-01-01

    Full Text Available A high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX and a secondary heat exchanger (SHX. Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C, high-pressure (7 MPa helium loop thermally integrated with a molten fluoride salt (KF-ZrF4 flow loop operating at low pressure (0.2 MPa, at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift in measuring operational data for extended periods of times, as data collected will be

  2. Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Barnett, P.C. [Canadian General Electric (Canada)

    1968-07-15

    The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

  3. Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

    Energy Technology Data Exchange (ETDEWEB)

    Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

    1982-01-01

    The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling.

  4. The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

    Directory of Open Access Journals (Sweden)

    Mario Carelli

    2009-01-01

    Full Text Available IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.

  5. Review of nuclear data improvement needs for nuclear radiation measurement techniques used at the CEA experimental reactor facilities

    Directory of Open Access Journals (Sweden)

    Destouches Christophe

    2016-01-01

    Full Text Available The constant improvement of the neutron and gamma calculation codes used in experimental nuclear reactors goes hand in hand with that of the associated nuclear data libraries. The validation of these calculation schemes always requires the confrontation with integral experiments performed in experimental reactors to be completed. Nuclear data of interest, straight as cross sections, or elaborated ones such as reactivity, are always derived from a reaction rate measurement which is the only measurable parameter in a nuclear sensor. So, in order to derive physical parameters from the electric signal of the sensor, one needs specific nuclear data libraries. This paper presents successively the main features of the measurement techniques used in the CEA experimental reactor facilities for the on-line and offline neutron/gamma flux characterizations: reactor dosimetry, neutron flux measurements with miniature fission chambers and Self Power Neutron Detector (SPND and gamma flux measurements with chamber ionization and TLD. For each technique, the nuclear data necessary for their interpretation will be presented, the main identified needs for improvement identified and an analysis of their impact on the quality of the measurement. Finally, a synthesis of the study will be done.

  6. Monitoring of dry anaerobic fermentation in experimental facility with use of biofilm reactor

    Directory of Open Access Journals (Sweden)

    Milan Šinkora

    2011-01-01

    -called biofilm reactor. An external reactor with a cultivated bacterial biofilm on an immovable carrier with the percolate flowing through it has been constructed in laboratory conditions for this purpose. The choice of suitable percolate strategy (this means the frequency of sprinkling and the amount of percolate directly influences the process of anaerobic fermentation.

  7. Large-scale experimental facility for emergency condition investigation of a new generation NPP WWER-640 reactor with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Aniskevich, Y.N.; Vasilenko, V.A.; Zasukha, V.K.; Migrov, Y.A.; Khabensky, V.B. [Research Inst. of Technology NITI (Russian Federation)

    1997-12-31

    The creation of the large-scale integral experimental facility (KMS) is specified by the programme of the experimental investigations to justify the engineering decisions on the safety of the design of the new generation NPP with the reactor WWER-640. The construction of KMS in a full volume will allow to conduct experimental investigations of all physical phenomena and processes, practically, occurring during the accidents on the NPPs with the reactor of WWER type and including the heat - mass exchange processes with low rates of the coolant, which is typical during the utilization of the passive safety systems, process during the accidents with a large leak, and also the complex intercommunicated processes in the reactor unit, passive safety systems and in the containment with the condition of long-term heat removal to the final absorber. KMS is being constructed at the Research Institute of Technology (NITI), Sosnovy Bor, Leningrad region, Russia. (orig.). 5 refs.

  8. Experimental Breeder Reactor I Preservation Plan

    Energy Technology Data Exchange (ETDEWEB)

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  9. Neutron dosimetry at the reactor facility VENUS

    Energy Technology Data Exchange (ETDEWEB)

    Deboodt, P.; Vermeersch, F.; Vanhavere, F.; Minsart, G. [SCK-CEN, Belgian Nuclear Research Centre, Mol (Belgium)

    1997-09-01

    The reactor VENUS is a zero-power research reactor mainly devoted to studies on light water fuels. The need for undertaking a neutron spectrometric and dosimetric study became apparent when locally high neutron dose rates were measured. The spectrometric study is based on two approaches. The first is an experimental one in which the neutron spectrum was measured at three positions around the facility. The second is a theoretical one in which a numerical modelling of the neutron transport at the reactor site was performed in order to determine neutron spectra and fluence rates at different positions around the site. The measured and calculated spectra are interpreted in terms of the responses of different individual and environmental dosemeters. These responses are confronted with the in situ measurements. The impact of the ICRP 60 recommendations on the determined dose rates is also studied. (author).

  10. Transonic Experimental Research Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Transonic Experimental Research Facility evaluates aerodynamics and fluid dynamics of projectiles, smart munitions systems, and sub-munitions dispensing systems;...

  11. Study of fast reactor safety test facilities. Preliminary report

    Energy Technology Data Exchange (ETDEWEB)

    Bell, G.I.; Boudreau, J.E.; McLaughlin, T.; Palmer, R.G.; Starkovich, V.; Stein, W.E.; Stevenson, M.G.; Yarnell, Y.L.

    1975-05-01

    Included are sections dealing with the following topics: (1) perspective and philosophy of fast reactor safety analysis; (2) status of accident analysis and experimental needs; (3) experiment and facility definitions; (4) existing in-pile facilities; (5) new facility options; and (6) data acquisition methods. (DG)

  12. Development of BWR regional stability experimental facility SIRIUS-F, which simulates thermal-hydraulics-neutronics coupling in reactor core, and stability evaluation of ABWR

    Energy Technology Data Exchange (ETDEWEB)

    Masahiro Furuya; Fumio Inada [Central Research Institute of Electric Power Industry (CRIEPI) 2-11-1 Iwado-kita, Komae, Tokyo 201-8511 (Japan); Takanori Fukahori [Global Nuclear Fuel Japan (GNF-J) 2-3-1 Uchikawa, Yokosuka, Kanagawa 239-0836 (Japan); Shinya Mizokami [Tokyo Electric Power Company (TEPCO) 1-1-3 Uchisaiwai-cho, Chiyoda, Tokyo 100-0011 (Japan)

    2005-07-01

    Full text of publication follows: The SIRIUS-F facility was designed and constructed for highly accurate simulation of channel, core-wide and regional instabilities of an ABWR. A real-time simulation is performed for the modal-point kinetics of reactor neutronics and fuel-rod conduction on the basis of a measured void fraction in a reactor core section of the facility. A noise analysis method was performed to calculate decay ratios from dominant poles of transfer function on the basis of the AR method by applying time series of a core inlet flow rate. By utilizing this method, one can estimate stability at any specific operating point online without assuming excess conservative conditions. Channel and regional stability experiments were conducted for a wide range of operating conditions including maximum power points along the minimum pump speed line and the natural circulation line of the ABWR. The decay ratios and the resonance frequencies are in good agreement with those from the design analysis code, ODYSY. The SIRIUS-F experimental results demonstrated stability characteristics such as a stabilizing effect of the power, and reviled a sufficiently large stability margin even under hypothetical conditions of power enlargement. (authors)

  13. Dosimetry and radiobiology at the new RA-3 reactor boron neutron capture therapy (BNCT) facility: Application to the treatment of experimental oral cancer

    Energy Technology Data Exchange (ETDEWEB)

    Pozzi, E. [Research and Production Reactors, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina)], E-mail: epozzi@cnea.gov.ar; Nigg, D.W. [Idaho National Laboratory, Idaho Falls (United States); Miller, M.; Thorp, S.I. [Instrumentation and Control Department, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Heber, E.M. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina); Zarza, L.; Estryk, G. [Research and Production Reactors, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Monti Hughes, A.; Molinari, A.J.; Garabalino, M. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina); Itoiz, M.E. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina); Department of Oral Pathology, Faculty of Dentistry, University of Buenos Aires (Argentina); Aromando, R.F. [Department of Oral Pathology, Faculty of Dentistry, University of Buenos Aires (Argentina); Quintana, J. [Research and Production Reactors, National Atomic Energy Commission, Ezeiza Atomic Center (Argentina); Trivillin, V.A.; Schwint, A.E. [Department of Radiobiology, National Atomic Energy Commission, Constituyentes Atomic Center (Argentina)

    2009-07-15

    The National Atomic Energy Commission of Argentina (CNEA) constructed a novel thermal neutron source for use in boron neutron capture therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The aim of the present study was to perform a dosimetric characterization of the facility and undertake radiobiological studies of BNCT in an experimental model of oral cancer in the hamster cheek pouch. The free-field thermal flux was 7.1x10{sup 9} n cm{sup -2} s{sup -1} and the fast neutron flux was 2.5x10{sup 6} n cm{sup -2} s{sup -1}, indicating a very well-thermalized neutron field with negligible fast neutron dose. For radiobiological studies it was necessary to shield the body of the hamster from the neutron flux while exposing the everted cheek pouch bearing the tumors. To that end we developed a lithium (enriched to 95% in {sup 6}Li) carbonate enclosure. Groups of tumor-bearing hamsters were submitted to BPA-BNCT, GB-10-BNCT, (GB-10+BPA)-BNCT or beam only treatments. Normal (non-cancerized) hamsters were treated similarly to evaluate normal tissue radiotoxicity. The total physical dose delivered to tumor with the BNCT treatments ranged from 6 to 8.5 Gy. Tumor control at 30 days ranged from 73% to 85%, with no normal tissue radiotoxicity. Significant but reversible mucositis in precancerous tissue surrounding tumors was associated to BPA-BNCT. The therapeutic success of different BNCT protocols in treating experimental oral cancer at this novel facility was unequivocally demonstrated.

  14. Dosimetry and radiobiology at the new RA-3 reactor boron neutron capture therapy (BNCT) facility: application to the treatment of experimental oral cancer.

    Science.gov (United States)

    Pozzi, E; Nigg, D W; Miller, M; Thorp, S I; Heber, E M; Zarza, L; Estryk, G; Monti Hughes, A; Molinari, A J; Garabalino, M; Itoiz, M E; Aromando, R F; Quintana, J; Trivillin, V A; Schwint, A E

    2009-07-01

    The National Atomic Energy Commission of Argentina (CNEA) constructed a novel thermal neutron source for use in boron neutron capture therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The aim of the present study was to perform a dosimetric characterization of the facility and undertake radiobiological studies of BNCT in an experimental model of oral cancer in the hamster cheek pouch. The free-field thermal flux was 7.1 x 10(9) n cm(-2)s(-1) and the fast neutron flux was 2.5 x 10(6) n cm(-2)s(-1), indicating a very well-thermalized neutron field with negligible fast neutron dose. For radiobiological studies it was necessary to shield the body of the hamster from the neutron flux while exposing the everted cheek pouch bearing the tumors. To that end we developed a lithium (enriched to 95% in (6)Li) carbonate enclosure. Groups of tumor-bearing hamsters were submitted to BPA-BNCT, GB-10-BNCT, (GB-10+BPA)-BNCT or beam only treatments. Normal (non-cancerized) hamsters were treated similarly to evaluate normal tissue radiotoxicity. The total physical dose delivered to tumor with the BNCT treatments ranged from 6 to 8.5 Gy. Tumor control at 30 days ranged from 73% to 85%, with no normal tissue radiotoxicity. Significant but reversible mucositis in precancerous tissue surrounding tumors was associated to BPA-BNCT. The therapeutic success of different BNCT protocols in treating experimental oral cancer at this novel facility was unequivocally demonstrated.

  15. Computational analysis of irradiation facilities at the JSI TRIGA reactor.

    Science.gov (United States)

    Snoj, Luka; Zerovnik, Gašper; Trkov, Andrej

    2012-03-01

    Characterization and optimization of irradiation facilities in a research reactor is important for optimal performance. Nowadays this is commonly done with advanced Monte Carlo neutron transport computer codes such as MCNP. However, the computational model in such calculations should be verified and validated with experiments. In the paper we describe the irradiation facilities at the JSI TRIGA reactor and demonstrate their computational characterization to support experimental campaigns by providing information on the characteristics of the irradiation facilities. Copyright © 2011 Elsevier Ltd. All rights reserved.

  16. The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

    Energy Technology Data Exchange (ETDEWEB)

    De Bruyn, D.; Engelen, J. [Belgian Nuclear Research Centre SCK CEN, Boeretang 200, 2400 Mol (Belgium); Ortega, A.; Aguado, M. P. [Empresarios Agrupados A.I.E., Magallanes 3, 28015 Madrid (Spain)

    2012-07-01

    MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of three years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)

  17. Facility for a Low Power Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chalker, R. G.

    1949-09-14

    Preliminary investigation indicates that a reactor facility with ample research provisions for use by University or other interested groups, featuring safety in design, can be economically constructed in the Los Angeles area. The complete installation, including an underground gas-tight reactor building, with associated storage and experiment assembly building, administration offices, two general laboratory buildings, hot latoratory and lodge, can be constructed for approxinately $1,500,000. This does not include the cost of the reactor itself or of its auxiliary equipment,

  18. Review of Transient Testing of Fast Reactor Fuels in the Transient REActor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, C.; Wachs, D.; Carmack, J.; Woolstenhulme, N.

    2017-01-01

    The restart of the Transient REActor Test (TREAT) facility provides a unique opportunity to engage the fast reactor fuels community to reinitiate in-pile experimental safety studies. Historically, the TREAT facility played a critical role in characterizing the behavior of both metal and oxide fast reactor fuels under off-normal conditions, irradiating hundreds of fuel pins to support fast reactor fuel development programs. The resulting test data has provided validation for a multitude of fuel performance and severe accident analysis computer codes. This paper will provide a review of the historical database of TREAT experiments including experiment design, instrumentation, test objectives, and salient findings. Additionally, the paper will provide an introduction to the current and future experiment plans of the U.S. transient testing program at TREAT.

  19. Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2017-05-15

    Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

  20. THE WHITE SANDS MISSILE RANGE PULSED REACTOR FACILITY, MAY 1963

    Energy Technology Data Exchange (ETDEWEB)

    Long, Robert L.; Boor, R.A.; Cole, W.M.; Elder, G.E.

    1963-05-15

    A brief statement of the mission of the White Sands Missile Range Nuclear Effects Laboratory is given. The new Nuclear Effects Laboratory Facility is described. This facility consists of two buildings-a laboratory and a reactor building. The White Sands Missile Range bare critical assembly, designated as the MoLLY-G, is described. The MoLLY-G, an unreflected, unmoderated right circular cylinder of uranium-molybdenum alloy designed for pulsed operation, will have a maximum burst capability of approximately 2 x 10/sup 17/ fissions with a burst width of 50 microseconds. The reactor construction and operating procedures are described. As designed, the MoLLY-G will provide an intense source of pulsed neutron and gamma radiation for a great variety of experimental and test arrangements. (auth)

  1. Feasibility of processing the experimental breeder reactor-II driver fuel from the Idaho National Laboratory through Savannah River Site's H-Canyon facility

    Energy Technology Data Exchange (ETDEWEB)

    Magoulas, V. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-28

    Savannah River National Laboratory (SRNL) was requested to evaluate the potential to receive and process the Idaho National Laboratory (INL) uranium (U) recovered from the Experimental Breeder Reactor II (EBR-II) driver fuel through the Savannah River Site’s (SRS) H-Canyon as a way to disposition the material. INL recovers the uranium from the sodium bonded metallic fuel irradiated in the EBR-II reactor using an electrorefining process. There were two compositions of EBR-II driver fuel. The early generation fuel was U-5Fs, which consisted of 95% U metal alloyed with 5% noble metal elements “fissium” (2.5% molybdenum, 2.0% ruthenium, 0.3% rhodium, 0.1% palladium, and 0.1% zirconium), while the later generation was U-10Zr which was 90% U metal alloyed with 10% zirconium. A potential concern during the H-Canyon nitric acid dissolution process of the U metal containing zirconium (Zr) is the explosive behavior that has been reported for alloys of these materials. For this reason, this evaluation was focused on the ability to process the lower Zr content materials, the U-5Fs material.

  2. Multi-Directional Experimental Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The ATLSS Multi-directional Experimental Laboratory was constructed in 1987 under funding from the National Science Foundation to be a major facility for large-scale...

  3. Commercial Light Water Reactor Tritium Extraction Facility

    Energy Technology Data Exchange (ETDEWEB)

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  4. Sandia Pulsed Reactor Facility (SPRF) calculator-assisted pulse analysis and display system

    Energy Technology Data Exchange (ETDEWEB)

    Estes, B.F.; Berry, D.T.

    1980-02-01

    Two solid-metal fast burst type reactors (SPR II and SPR III) are operated at the Sandia Pulsed Reactor Facility. Since startup of the reactors, oscilloscope traces have been used to record (by camera) the pulse (power) shape while log N systems have measured initial reactor period. Virtually no other pulse information is available. A decision was made to build a system that could collect the basic input data available from the reactor - fission chambers, photodiodes, and thermocouples - condition the signals and output the various parameters such as power, energy, temperature, period and lifetime on hard copy that would provide a record for operations personnel as well as the experimenter. Because the reactors operate in short time frames - pulse operation - it is convenient to utilize the classical Nordheim-Fuchs approximation of the diffusion equation to describe reactor behavior. This report describes the work performed to date in developing the calculator system and analytical models for computing the desired parameters.

  5. The Automation of the Transonic Experimental Facility (TEF) and the Aerodynamic Experimental Facility (AEF)

    Science.gov (United States)

    2015-10-01

    Aerodynamic Experimental Facility (AEF) by Charith R Ranawake Approved for public release; distribution is unlimited...Laboratory The Automation of the Transonic Experimental Facility (TEF) and the Aerodynamic Experimental Facility (AEF) by Charith R Ranawake Weapons...To) 05/2015–08/2015 4. TITLE AND SUBTITLE The Automation of the Transonic Experimental Facility (TEF) and the Aerodynamic Experimental Facility

  6. Seclazone Reactor Modeling And Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Osinga, T. [ETH-Zuerich (Switzerland); Olalde, G. [CNRS Odeillo (France); Steinfeld, A. [PSI and ETHZ (Switzerland)

    2005-03-01

    A numerical model is formulated for the SOLZINC solar chemical reactor for the production of Zn by carbothermal reduction of ZnO. The model involves solving, by the finite-volume technique, a 1D unsteady state energy equation that couples heat transfer to the chemical kinetics for a shrinking packed bed exposed to thermal radiation. Validation is accomplished by comparison with experimentally measured temperature profiles and Zn production rates as a function of time, obtained for a 5-kW solar reactor tested at PSI's solar furnace. (author)

  7. Program for studying fundamental interactions at the PIK reactor facilities

    Science.gov (United States)

    Serebrov, A. P.; Vassiljev, A. V.; Varlamov, V. E.; Geltenbort, P.; Gridnev, K. A.; Dmitriev, S. P.; Dovator, N. A.; Egorov, A. I.; Ezhov, V. F.; Zherebtsov, O. M.; Zinoviev, V. G.; Ivochkin, V. G.; Ivanov, S. N.; Ivanov, S. A.; Kolomensky, E. A.; Konoplev, K. A.; Krasnoschekova, I. A.; Lasakov, M. S.; Lyamkin, V. A.; Martemyanov, V. P.; Murashkin, A. N.; Neustroev, P. V.; Onegin, M. S.; Petelin, A. L.; Pirozhkov, A. N.; Polyushkin, A. O.; Prudnikov, D. V.; Ryabov, V. L.; Samoylov, R. M.; Sbitnev, S. V.; Fomin, A. K.; Fomichev, A. V.; Zimmer, O.; Cherniy, A. V.; Shoka, I. V.

    2016-05-01

    A research program aimed at studying fundamental interactions by means of ultracold and polarized cold neutrons at the GEK-4-4' channel of the PIK reactor is presented. The apparatus to be used includes a source of cold neutrons in the heavy-water reflector of the reactor, a source of ultracold neutrons based on superfluid helium and installed in a cold-neutron beam extracted from the GEK-4 channel, and a number of experimental facilities in neutron beams. An experiment devoted to searches for the neutron electric dipole moment and an experiment aimed at a measurement the neutron lifetime with the aid of a large gravitational trap are planned to be performed in a beam of ultracold neutrons. An experiment devoted to measuring neutron-decay asymmetries with the aid of a superconducting solenoid is planned in a beam of cold polarized neutrons from the GEK-4' channel. The second ultracold-neutron source and an experiment aimed at measuring the neutron lifetime with the aid of a magnetic trap are planned in the neutron-guide system of the GEK-3 channel. In the realms of neutrino physics, an experiment intended for sterile-neutrino searches is designed. The state of affairs around the preparation of the experimental equipment for this program is discussed.

  8. Analyses of the OSU-MASLWR Experimental Test Facility

    Directory of Open Access Journals (Sweden)

    F. Mascari

    2012-01-01

    Full Text Available Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR design, a small modular pressurized water reactor (PWR, relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characterizing the tests already performed, the potential transients that could be investigated in the facility, and to describe the current IAEA International Collaborative Standard Problem that is being hosted at OSU and the experimental data will be collected at the OSU-MASLWR test facility. A summary of the best estimate thermal hydraulic system code analyses, already performed, to analyze the codes capability in predicting the phenomena typical of the MASLWR prototype, thermal hydraulically characterized in the OSU-MASLWR facility, is presented as well.

  9. Preconceptual design of the new production reactor circulator test facility

    Energy Technology Data Exchange (ETDEWEB)

    Thurston, G.

    1990-06-01

    This report presents the results of a study of a new circulator test facility for the New Production Reactor Modular High-Temperature Gas-Cooled Reactor. The report addresses the preconceptual design of a stand-alone test facility with all the required equipment to test the Main Circulator/shutoff valve and Shutdown Cooling Circulator/shutoff valve. Each type of circulator will be tested in its own full flow, full power helium test loop. Testing will cover the entire operating range of each unit. The loop will include a test vessel, in which the circulator/valve will be mounted, and external piping. The external flow piping will include a throttle valve, flowmeter, and heat exchanger. Subsystems will include helium handling, helium purification, and cooling water. A computer-based data acquisition and control system will be provided. The estimated costs for the design and construction of this facility are included. 2 refs., 15 figs.

  10. Calculation and comparisons with measurement of fast neutron fluxes in the material testing facilities of the NRU research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leung, T.C. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    The NRU reactor at Chalk River provides three irradiation facilities to study the effects of fast neutrons (E> 1 MeV) on reactor materials for assessing material damage and deformation. The facilities comprise two types of fast neutron rods (Mark 4 and Mark 7), and a Material Test Bundle (MTB) irradiated in a loop site. This paper describes the neutronic simulation of these testing facilities using the WIMS-AECL and TRIAD codes, and comparisons with the fast neutron flux measurements using iron-wire activation techniques. It also provides comparisons of flux levels, neutron spectra, and size limitations of the experimental cavities between these test facilities. (author)

  11. Heat barrier for use in a nuclear reactor facility

    Science.gov (United States)

    Keegan, Charles P.

    1988-01-01

    A thermal barrier for use in a nuclear reactor facility is disclosed herein. Generally, the thermal barrier comprises a flexible, heat-resistant web mounted over the annular space between the reactor vessel and the guard vessel in order to prevent convection currents generated in the nitrogen atmosphere in this space from entering the relatively cooler atmosphere of the reactor cavity which surrounds these vessels. Preferably, the flexible web includes a blanket of heat-insulating material formed from fibers of a refractory material, such as alumina and silica, sandwiched between a heat-resistant, metallic cloth made from stainless steel wire. In use, the web is mounted between the upper edges of the guard vessel and the flange of a sealing ring which surrounds the reactor vessel with a sufficient enough slack to avoid being pulled taut as a result of thermal differential expansion between the two vessels. The flexible web replaces the rigid and relatively complicated structures employed in the prior art for insulating the reactor cavity from the convection currents generated between the reactor vessel and the guard vessel.

  12. Integral Test Facility PKL: Experimental PWR Accident Investigation

    Directory of Open Access Journals (Sweden)

    Klaus Umminger

    2012-01-01

    Full Text Available Investigations of the thermal-hydraulic behavior of pressurized water reactors under accident conditions have been carried out in the PKL test facility at AREVA NP in Erlangen, Germany for many years. The PKL facility models the entire primary side and significant parts of the secondary side of a pressurized water reactor (PWR at a height scale of 1 : 1. Volumes, power ratings and mass flows are scaled with a ratio of 1 : 145. The experimental facility consists of 4 primary loops with circulation pumps and steam generators (SGs arranged symmetrically around the reactor pressure vessel (RPV. The investigations carried out encompass a very broad spectrum from accident scenario simulations with large, medium, and small breaks, over the investigation of shutdown procedures after a wide variety of accidents, to the systematic investigation of complex thermal-hydraulic phenomena. This paper presents a survey of test objectives and programs carried out to date. It also describes the test facility in its present state. Some important results obtained over the years with focus on investigations carried out since the beginning of the international cooperation are exemplarily discussed.

  13. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. [ed.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG&G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  14. Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ryskamp, J.M. (ed.); Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

    1992-07-01

    This report describes the EG G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options.

  15. Monitoring system for an experimental facility using GMDH methodology

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Iraci Martinez; Moraes, Davi Almeida [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Bueno, Elaine Inacio, E-mail: martinez@ipen.br, E-mail: dmoraes@dk8.com.br, E-mail: ebueno@ifsp.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia de Sao Paulo (IFSP), São Paulo, SP (Brazil)

    2017-07-01

    This work presents a Monitoring System developed based on the GMDH - Group Method of Data Handling methodology to be used in an Experimental Test Facility. GMDH is a combinatorial multi-layer algorithm in which a network of layers and nodes is generated using a number of inputs from the data stream being evaluated. The GMDH network topology has been traditionally determined using a layer by layer pruning process based on a pre-selected criterion of what constitutes the best nodes at each level. The traditional GMDH method is based on an underlying assumption that the data can be modeled by using an approximation of the Volterra Series or Kolmorgorov-Gabor polynomial. The Fault Test Experimental Facility was designed to simulate a PWR nuclear power plant and is composed by elements that correspond to the pressure vessel, steam generator, pumps of the primary and secondary reactor loops. The nuclear reactor core is represented by an electrical heater with different values of power. The experimental plant will be fully instrumented with sensors and actuators, and the data acquisition system will be constructed in order to enable the details of the temporal analysis of process variables. The Fault Test Experimental Facility can be operated to generate normal and fault data. These failures can be added initially with small magnitude, and their magnitude being increasing gradually in a controlled way. The database will interface with the plant supervisory system SCADA (Supervisory Control and Data Acquisition) that provides the data through standard interface. (author)

  16. Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

    Science.gov (United States)

    Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

    2017-12-01

    The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

  17. Proposed design for the PGAA facility at the TRIGA IPR-R1 research reactor.

    Science.gov (United States)

    Guerra, Bruno T; Jacimovic, Radojko; Menezes, Maria Angela Bc; Leal, Alexandre S

    2013-01-01

    This work presents an initial proposed design of a Prompt Gamma Activation Analysis (PGAA) facility to be installed at the TRIGA IPR-R1, a 60 years old research reactor of the Centre of Development of Nuclear Technology (CDTN) in Brazil. The basic characteristics of the facility and the results of the neutron flux are presented and discussed. The proposed design is based on a quasi vertical tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below the room's level where shall be located the rack containing the set sample/detector/shielding. The evaluation of the thermal and epithermal neutron flux in the sample position was done considering the experimental data obtained from a vertical neutron guide, already existent in the reactor, and the simulated model for the facility. The experimental determination of the neutron flux was obtained through the standard procedure of using Au monitors in different positions of the vertical tube. In order to validate both, this experiment and calculations of the simulated model, the flux was also determined in different positions in the core used for sample irradiation. The model of the system was developed using the Monte Carlo code MCNP5. The preliminary results suggest the possibility of obtaining a beam with minimum thermal flux of magnitude 10(6) cm(-2) s(-1), which confirm the technical feasibility of the installation of PGAA at the TRIGA IPR-R1 reactor. This beam would open new possibilities for enhancing the applications using the reactor.

  18. Final report of the HFIR (High Flux Isotope Reactor) irradiation facilities improvement project

    Energy Technology Data Exchange (ETDEWEB)

    Montgomery, B.H.; Thoms, K.R.; West, C.D.

    1987-09-01

    The High-Flux Isotope Reactor (HFIR) has outstanding neutronics characteristics for materials irradiation, but some relatively minor aspects of its mechanical design severely limited its usefulness for that purpose. In particular, though the flux trap region in the center of the annular fuel elements has a very high neutron flux, it had no provision for instrumentation access to irradiation capsules. The irradiation positions in the beryllium reflector outside the fuel elements also have a high flux; however, although instrumented, they were too small and too few to replace the facilities of a materials testing reactor. To address these drawbacks, the HFIR Irradiation Facilities Improvement Project consisted of modifications to the reactor vessel cover, internal structures, and reflector. Two instrumented facilities were provided in the flux trap region, and the number of materials irradiation positions in the removable beryllium (RB) was increased from four to eight, each with almost twice the available experimental space of the previous ones. The instrumented target facilities were completed in August 1986, and the RB facilities were completed in June 1987.

  19. 1996 environmental monitoring report for the Naval Reactors Facility

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-31

    The results of the radiological and nonradiological environmental monitoring programs for 1996 at the Naval Reactors Facility (NRF) are presented in this report. The NRF is located on the Idaho National Engineering and Environmental Laboratory and contains three naval reactor prototypes and the Expended Core Facility, which examines developmental nuclear fuel material samples, spent naval fuel, and irradiated reactor plant components/materials. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE).

  20. 77 FR 26321 - Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112

    Science.gov (United States)

    2012-05-03

    ... COMMISSION Reed College, Reed Research Nuclear Reactor, Renewed Facility Operating License No. R-112 AGENCY..., Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Rockville, MD 20852. Telephone..., Research and Test Reactors Licensing Branch, Division of Policy and Rulemaking, Office of Nuclear Reactor...

  1. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  2. The new neutron imaging facility at TRIGA reactor in Morocco

    Energy Technology Data Exchange (ETDEWEB)

    Ouardi, A.; Alami, R.; Bensitel, A. [Centre National de l' Energie des Science et des Techniques Nucleaires, PB.1382 R.P 10001 Rabat (Morocco)

    2011-07-01

    A new neutron imaging facility is currently developed around 2 MW TRIGA MARK-II reactor at Maamora Nuclear research centre (CENM). Neutron imaging combined to X-ray or gamma radiography offers the opportunity to extend Non Destructive Testing (NDT) activities DT in Morocco to new fields of applications such as space and aircraft Moroccan industry, mining, wood industry and Archeology. The facility is planed to be completed in the end of 2011. In order to reduce the gamma-ray content in the neutron beam, the reactor tangential channel is selected. For power of 2 MW, the corresponding thermal neutron flux at the inlet of the tangential channel is around 1.10{sup 13}ncm{sup 2}/s. The facility will be based on a conical neutron collimator with a flight tube of 8m and offers three circular diaphragms with diameters of 1cm, 2 cm and 4 cm corresponding to L/D-ratio varying between 200 and 600. The holes will be housed in the primary shutter. These diaphragms' sizes allow to perform neutron radiography with high resolution (L/D = 600) and high speed (L/D= 200). Monte Carlo calculations by a fully 3D numerical code GEANT4 are used to optimize the whole neutron beam line and to reach a shorten distance between the source and detector and reduce as possible the exposure time. (author)

  3. The CABRI fast neutron Hodoscope: Renovation, qualification program and first results following the experimental reactor restart

    Science.gov (United States)

    Chevalier, V.; Mirotta, S.; Guillot, J.; Biard, B.

    2018-01-01

    The CABRI experimental pulse reactor, located at the Cadarache nuclear research center, southern France, is devoted to the study of Reactivity Initiated Accidents (RIA). For the purpose of the CABRI International Program (CIP), managed and funded by IRSN, in the framework of an OECD/NEA agreement, a huge renovation of the facility has been conducted since 2003. The Cabri Water Loop was then installed to ensure prototypical Pressurized Water Reactor (PWR) conditions for testing irradiated fuel rods. The hodoscope installed in the CABRI reactor is a unique online fuel motion monitoring system, operated by IRSN and dedicated to the measurement of the fast neutrons emitted by the tested rod during the power pulse. It is one of the distinctive features of the CABRI reactor facility, which is operated by CEA. The system is able to determine the fuel motion, if any, with a time resolution of 1 ms and a spatial resolution of 3 mm. The hodoscope equipment has been upgraded as well during the CABRI facility renovation. This paper presents the main outcomes achieved with the hodoscope since October 2015, date of the first criticality of the CABRI reactor in its new Cabri Water Loop configuration. Results obtained during reactor commissioning phase functioning, either in steady-state mode (at low and high power, up to 23 MW) or in transient mode (start-up, possibly beyond 20 GW), are discussed.

  4. The TRIGA Reactor Facility at the Armed Forces Radiobiology Research Institute: A Simplified Technical Description.

    Science.gov (United States)

    1986-05-01

    AD-AiS68 238 THE TRIGA REACTOR FACILITY AT THE ARMED FORCESI! RADIOBIOLOGY RESEARCH INST..(U) ARMED FORCES RADIOBIOLOGY RESEARCH INST BETHESDA NO...medium-power exposure. The reactor is also used to train military personnel in reactor operations., The AFRRI TRIGA Mark-F reactor facility is within the...AFRRI complex on the grounds ’of the Naval Medical Command National Capital Region, in Bethesda, Maryland.> TRIGA is an acronym for Training, Research

  5. CERN inaugurates its latest experimental facility : ISOLDE

    CERN Document Server

    Rubbia, C; Detraz, C; CERN. Geneva

    1992-05-25

    Leading physicists from all over Europe came to CERN on 26 May to celebrate the inauguration of CERN'S latest experimental facility, ISOLDE (Isotope Separator On-Line) at the Proton Synchrotron Booster. A ceremony was held in the new ISOLDE experimental hall where the participants were welcomed by Prof. C. Rubbia, Director General of CERN. Prof. B. Jonson, of Chalmers University in Göteborg, explained the physics potential of the new installation and Prof. C. Detraz, Director of the Institut National de Physique Nucléaire et Physique des Particules (IN2P3), stressed the importance of the ISOLDE facility for European Nuclar Physics.

  6. Extraction of pure thermal neutron beam for the proposed PGNAA facility at the TRIGA research reactor of AERE, Savar, Bangladesh

    Science.gov (United States)

    Alam, Sabina; Zaman, M. A.; Islam, S. M. A.; Ahsan, M. H.

    1993-10-01

    A study on collimators and filters for the design of a spectrometer for prompt gamma neutron activation analysis (PGNAA) at one of the radial beamports of the TRIGA Mark II reactor at AERE, Savar has been carried out. On the basis of this study a collimator and a filter have been designed for the proposed PGNAA facility. Calculations have been done for measuring neutron flux at various positions of the core of the reactor using the computer code TRIGAP. Gamma dose in the core of the reactor has also been measured experimentally using TLD technique in the present work.

  7. Integral Reactor Containment Condensation Model and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Qiao [Oregon State Univ., Corvallis, OR (United States); Corradini, Michael [Univ. of Wisconsin, Madison, WI (United States)

    2016-05-02

    This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). In the three years of investigation duration, following the original proposal, the planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental conditions, and predicted the prototypic containment performance under accidental conditions (UW-Madison). A film flow model was developed for the scaling analysis, and the results suggest that the 1/3 scaled test facility covers large portion of laminar film flow, leading to a lower average heat transfer coefficient comparing to the prototypic value. Although it is conservative in reactor safety analysis, the significant reduction of heat transfer coefficient (50%) could under estimate the prototypic condensation heat transfer rate, resulting in inaccurate prediction of the decay heat removal capability. Further investigation is thus needed to quantify the scaling distortion for safety analysis code validation. Experimental investigations were performed in the existing MASLWR test facility at OrST with minor modifications. A total of 13 containment condensation tests were conducted for pressure

  8. The D&D of the Experimental Boiling Water Reactor (EBWR)

    Energy Technology Data Exchange (ETDEWEB)

    Fellhauer, C.R.; Boling, L.E.; Yule, T.J.; Bhattacharyya, S.K.

    1996-03-01

    Argonne National Laboratory has completed the D&D of the Experimental Boiling Water Reactor. The Project consisted of decontaminating and for packaging as radioactive waste the reactor vessel and internals, contaminated piping systems, miscellaneous tanks, pumps, and associated equipment. The D&D work involved dismantling process equipment and associated plumbing, ductwork drain lines, etc., performing size reduction of reactor vessel internals in the fuel pool, packaging and manifesting all radioactive and mixed waste, and performing a thorough survey of the facility after the removal of activated and contaminated material. Non-radioactive waste was disposed of in the ANL-E landfill or recycled. In January 1996 the EBWR facility was formally decommissioned and transferred from EM-40 to EM-30. This paper will discuss the details of this ten year effort.

  9. Experimental Fuels Facility Re-categorization Based on Facility Segmentation

    Energy Technology Data Exchange (ETDEWEB)

    Reiss, Troy P.; Andrus, Jason

    2016-07-01

    The Experimental Fuels Facility (EFF) (MFC-794) at the Materials and Fuels Complex (MFC) located on the Idaho National Laboratory (INL) Site was originally constructed to provide controlled-access, indoor storage for radiological contaminated equipment. Use of the facility was expanded to provide a controlled environment for repairing contaminated equipment and characterizing, repackaging, and treating waste. The EFF facility is also used for research and development services, including fuel fabrication. EFF was originally categorized as a LTHC-3 radiological facility based on facility operations and facility radiological inventories. Newly planned program activities identified the need to receive quantities of fissionable materials in excess of the single parameter subcritical limit in ANSI/ANS-8.1, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors” (identified as “criticality list” quantities in DOE-STD-1027-92, “Hazard Categorization and Accident Analysis Techniques for Compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports,” Attachment 1, Table A.1). Since the proposed inventory of fissionable materials inside EFF may be greater than the single parameter sub-critical limit of 700 g of U-235 equivalent, the initial re-categorization is Hazard Category (HC) 2 based upon a potential criticality hazard. This paper details the facility hazard categorization performed for the EFF. The categorization was necessary to determine (a) the need for further safety analysis in accordance with LWP-10802, “INL Facility Categorization,” and (b) compliance with 10 Code of Federal Regulations (CFR) 830, Subpart B, “Safety Basis Requirements.” Based on the segmentation argument presented in this paper, the final hazard categorization for the facility is LTHC-3. Department of Energy Idaho (DOE-ID) approval of the final hazard categorization determined by this hazard assessment document (HAD) was required per the

  10. Plasma heating systems planned for the Argonne experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bertoncini, P.; Brooks, J.; Fasolo, J.; Mills, F.; Moretti, A.; Norem, J.

    1976-01-01

    A scoping study and conceptual design of a tokamak experimental power reactor (TEPR) have been completed. The design objectives of the TEPR are to operate for ten years at or near electrical power breakeven conditions with a duty factor of greater than or equal to 50 percent and to demonstrate the feasibility of tokamak fusion power reactor techniques. These objectives can be met by a design which has a major radius of 6.25 m and a plasma radius of 2.1 m. Parameters for this reactor are listed, and a diagram is given. This paper will describe TEPR plasma heating systems. Neutral beam heating and rf heating are described.

  11. Upgrading program of the experimental fast reactor Joyo

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, A.; Yogo, S. [Japan Nuclear Cycle Development Institute, Iibaraki-Ken (Japan)

    2001-07-01

    The experimental fast reactor Joyo finished its operation as an irradiation core in June, 2000. Throughout the operation of MK-I (breeder core) and MK-II (irradiation core), the net operation time has exceeded 60,000 hours. During these operations there were no fuel failures or serious plant problems. The MK-III modification program will improve irradiation capability to demonstrate advanced technologies for commercial Fast Breeder Reactor (FBR). When the MK-III core is started, it will support irradiation tests in feasibility studies for fast reactor and related fuel cycle research and development in Japan. (authors)

  12. Naval Reactors Facility Environmental Monitoring Report, Calendar Year 2003

    Energy Technology Data Exchange (ETDEWEB)

    None

    2003-12-31

    The results of the radiological and nonradiological environmental monitoring programs for 2003 at the Naval Reactors Facility are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency and the U.S. Department of Energy.

  13. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Lewis, M R

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  14. Naval Reactors Facility environmental monitoring report, calendar year 2000

    Energy Technology Data Exchange (ETDEWEB)

    None

    2001-12-01

    The results of the radiological and nonradiological environmental monitoring programs for 2000 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency (EPA) and the U.S. Department of Energy (DOE).

  15. 1997 environmental monitoring report for the Naval Reactors Facility

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    The results of the radiological and nonradiological environmental monitoring programs for 1997 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE).

  16. Naval Reactors Facility environmental monitoring report, calendar year 2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-12-31

    The results of the radiological and nonradiological environmental monitoring programs for 2001 at the Naval Reactors Facility are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U. S. Environmental Protection Agency and the U. S. Department of Energy.

  17. Naval Reactors Facility environmental monitoring report, calendar year 1999

    Energy Technology Data Exchange (ETDEWEB)

    None

    2000-12-01

    The results of the radiological and nonradiological environmental monitoring programs for 1999 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency (EPA) and the U.S. Department of Energy (DOE).

  18. 1993 environmental monitoring report for the naval reactors facility

    Energy Technology Data Exchange (ETDEWEB)

    1994-11-01

    The results of the radiological and nonradiological environmental monitoring programs for 1993 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE).

  19. Investigation of mixing chamber for experimental FGD reactor

    Directory of Open Access Journals (Sweden)

    Novosád Jan

    2017-01-01

    Full Text Available This article deals with numerical investigation of flow and mixing of air and sulphur dioxide SO2 in designated mixing chamber. The mixing chamber is a part of experimental laboratory reactor designed for simulating the flue gas desulfurization (FGD process. Aim of this work is the numerical investigation of effect of different mixing chamber geometries to mixture composition, especially to mass fraction of sulphur dioxide. Using of similar concentration of sulphur dioxide in the experimental reactor as in the real process is necessary to be able to make additional research. Conclusion describes the effect of different geometries of mixing chamber to mixing. The aim of this work is to develop perfectly works mixing chamber, which will be manufactured and then implemented into experimental FGD reactor. The results will be validated by experiment after the mixing chamber will be manufactured.

  20. Geothermal Loop Experimental Facility. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1980-04-01

    Research at the Geothermal Loop Experimental Facility was successfully concluded in September 1979. In 13,000 hours of operation over a three and one half year period, the nominal 10 megawatt electrical equivalent GLEF provided the opportunity to identify problems in working with highly saline geothermal fluids and to develop solutions that could be applied to a commercial geothermal power plant producing electricity. A seven and one half year period beginning in April 1972, with early well flow testing and ending in September 1979, with the completion of extensive facility and reservoir operations is covered. During this period, the facility was designed, constructed and operated in several configurations. A comprehensive reference document, addressing or referencing documentation of all the key areas investigated is presented.

  1. Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Krass, A.W.

    2005-12-19

    This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type reactor cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type reactor cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.

  2. MISTRAL: an experimental programme in the EOLE facility devoted to 100% MOX core physics

    Energy Technology Data Exchange (ETDEWEB)

    Cathalau, S.; Cabrillat, J.C.; Chauvin, J.P.; Finck, P.J.; Fougeras, P.; Flamenbaum, G. [CEA Centre d`Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France); Matsu-ura, H.; Ueji, M.; Yamamoto, T.

    1996-09-01

    An extensive experimental programme, MISTRAL, has been undertaken in the EOLE critical facility in order to measure the main neutronic parameters of 100% MOX loaded Advanced Light Water Reactor cores and to improve the core analysis methods for those cores. The experimental programme comprises four core configurations with high moderation ratio, including three homogeneous cores and one ALWR type mock-up core. This paper presents the core configurations, measured parameters and experimental techniques used during the programme. (author)

  3. Advanced Reactors-Intermediate Heat Exchanger (IHX) Coupling: Theoretical Modeling and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Utgikar, Vivek [Univ. of Idaho, Moscow, ID (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Christensen, Richard [The Ohio State Univ., Columbus, OH (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-12-29

    The overall goal of the research project was to model the behavior of the advanced reactorintermediate heat exchange system and to develop advanced control techniques for off-normal conditions. The specific objectives defined for the project were: 1. To develop the steady-state thermal hydraulic design of the intermediate heat exchanger (IHX); 2. To develop mathematical models to describe the advanced nuclear reactor-IHX-chemical process/power generation coupling during normal and off-normal operations, and to simulate models using multiphysics software; 3. To develop control strategies using genetic algorithm or neural network techniques and couple these techniques with the multiphysics software; 4. To validate the models experimentally The project objectives were accomplished by defining and executing four different tasks corresponding to these specific objectives. The first task involved selection of IHX candidates and developing steady state designs for those. The second task involved modeling of the transient and offnormal operation of the reactor-IHX system. The subsequent task dealt with the development of control strategies and involved algorithm development and simulation. The last task involved experimental validation of the thermal hydraulic performances of the two prototype heat exchangers designed and fabricated for the project at steady state and transient conditions to simulate the coupling of the reactor- IHX-process plant system. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed high-temperature molten salt facility.

  4. Experimental Facilities at the High Energy Frontier

    CERN Document Server

    Jenni, P.

    2016-01-01

    The main theme of the lectures covered the experimental work at hadron colliders, with a clear focus on the Large Hadron Collider (LHC) and on the roadmap that led finally to the discovery of the Higgs boson. The lectures were not a systematic course on machine and detector technologies, but rather tried to give a physics-motivated overview of many experimental aspects that were all relevant for making the discovery. The actual lectures covered a much broader scope than what is documented here in this write- up. The successful concepts for the experiments at the LHC have benefitted from the experience gained with previous generations of detectors at lower- energy machines. The lectures included also an outlook to the future experimental programme at the LHC, with its machine and experiments upgrades, as well as a short discussion of possible facilities at the high energy frontier beyond LHC.

  5. Training experience at Experimental Breeder Reactor II

    Energy Technology Data Exchange (ETDEWEB)

    Driscoll, J.W.; McCormick, R.P.; McCreery, H.I.

    1978-01-01

    The EBR-II Training Group develops, maintains,and oversees training programs and activities associated with the EBR-II Project. The group originally spent all its time on EBR-II plant-operations training, but has gradually spread its work into other areas. These other areas of training now include mechanical maintenance, fuel manufacturing facility, instrumentation and control, fissile fuel handling, and emergency activities. This report describes each of the programs and gives a statistical breakdown of the time spent by the Training Group for each program. The major training programs for the EBR-II Project are presented by multimedia methods at a pace controlled by the student. The Training Group has much experience in the use of audio-visual techniques and equipment, including video-tapes, 35 mm slides, Super 8 and 16 mm film, models, and filmstrips. The effectiveness of these techniques is evaluated in this report.

  6. Experimental studies of irradiated and hydrogen implantation damaged reactor steels

    Energy Technology Data Exchange (ETDEWEB)

    Slugeň, Vladimír, E-mail: vladimir.slugen@stuba.sk; Pecko, Stanislav; Sojak, Stanislav

    2016-01-15

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1–2 vacancies with relatively small contribution (with intensity on the level of 20–40 %) were observed in “as-received” steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2–3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

  7. Experimental studies of irradiated and hydrogen implantation damaged reactor steels

    Science.gov (United States)

    Slugeň, Vladimír; Pecko, Stanislav; Sojak, Stanislav

    2016-01-01

    Radiation degradation of nuclear materials can be experimentally simulated via ion implantation. In our case, German reactor pressure vessel (RPV) steels were studied by positron annihilation lifetime spectroscopy (PALS). This unique non-destructive method can be effectively applied for the evaluation of microstructural changes and for the analysis of degradation of reactor steels due to neutron irradiation and proton implantation. Studied specimens of German reactor pressure vessel steels are originally from CARINA/CARISMA program. Eight specimens were measured in as-received state and two specimens were irradiated by neutrons in German experimental reactor VAK (Versuchsatomkraftwerk Kahl) in the 1980s. One of the specimens which was in as-received and neutron irradiated condition was also used for simulation of neutron damage by hydrogen nuclei implantation. Defects with the size of about 1-2 vacancies with relatively small contribution (with intensity on the level of 20-40 %) were observed in "as-received" steels. A significant increase in the size of the induced defects due to neutron damage was observed in the irradiated specimens resulting in 2-3 vacancies. The size and intensity of defects reached a similar level as in the specimens irradiated in the nuclear reactor due to the implantation of hydrogen ions with energies of 100 keV (up to the depth <500 nm).

  8. Validation of High-Fidelity Reactor Physics Models for Support of the KJRR Experimental Campaign in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, David W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nielsen, Joseph W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Norman, Daren R. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-01

    The Korea Atomic Energy Research Institute is currently in the process of qualifying a Low-Enriched Uranium fuel element design for the new Ki-Jang Research Reactor (KJRR). As part of this effort, a prototype KJRR fuel element was irradiated for several operating cycles in the Northeast Flux Trap of the Advanced Test Reactor (ATR) at the Idaho National Laboratory. The KJRR fuel element contained a very large quantity of fissile material (618g 235U) in comparison with historical ATR experiment standards (<1g 235U), and its presence in the ATR flux trap was expected to create a neutronic configuration that would be well outside of the approved validation envelope for the reactor physics analysis methods used to support ATR operations. Accordingly it was necessary, prior to high-power irradiation of the KJRR fuel element in the ATR, to conduct an extensive set of new low-power physics measurements with the KJRR fuel element installed in the ATR Critical Facility (ATRC), a companion facility to the ATR that is located in an immediately adjacent building, sharing the same fuel handling and storage canal. The new measurements had the objective of expanding the validation envelope for the computational reactor physics tools used to support ATR operations and safety analysis to include the planned KJRR irradiation in the ATR and similar experiments that are anticipated in the future. The computational and experimental results demonstrated that the neutronic behavior of the KJRR fuel element in the ATRC is well-understood, both in terms of its general effects on core excess reactivity and fission power distributions, its effects on the calibration of the core lobe power measurement system, as well as in terms of its own internal fission rate distribution and total fission power per unit ATRC core power. Taken as a whole, these results have significantly extended the ATR physics validation envelope, thereby enabling an entire new class of irradiation experiments.

  9. Experimental estimation of moderator temperature coefficient of reactivity of the IPEN/MB-01 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Rubens C. da; Bitelli, Ulysses D.; Mura, Luiz Ernesto C., E-mail: rubensrcs@usp.br, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Universidade de Sao Paulo (PNV/POLI/USP), SP (Brazil). Arquitetura Naval e Departamento de Engenharia Oceanica; Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The aim of this article is to present the procedure for the experimental estimation of the Moderator Temperature Coefficient of Reactivity of the IPEN/MB-01 Research Reactor, a parameter that has an important role in the physics and the control operations of any reactor facility. At the experiment, the IPEN/MB-01 reactor went critical at the power of 1W (1% of its total power), and whose core configuration was 28 x 26 rectangular array of UO{sub 2} fuel rods, inside a light water (moderator) tank. In addition, there was a heavy water (D{sub 2}O) reflector installed in the West side of the core to obtain an adequate neutron reflection along the experiment. The moderator temperature was increased in steps of 4 °C, and the measurement of the mean moderator temperature was acquired using twelve calibrated thermocouples, placed around the reactor core. As a result, the mean value of -4.81 pcm/°C was obtained for such coefficient. The curves of ρ(T) (Reactivity x Temperature) and α{sup M}{sub T}(T)(Moderator Temperature Coefficient of Reactivity x Temperature) were developed using data from an experimental measurement of the integral reactivity curves through the Stable Period and Inverse Kinetics Methods, that was carried out at the reactor with the same core configuration. Such curves were compared and showed a very similar behavior between them. (author)

  10. Compressed Gas Safety for Experimental Fusion Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Lee C. Cadwallader

    2004-09-01

    Experimental fusion facilities present a variety of hazards to the operators and staff. There are unique or specialized hazards, including magnetic fields, cryogens, radio frequency emissions, and vacuum reservoirs. There are also more general industrial hazards, such as a wide variety of electrical power, pressurized air, and cooling water systems in use, there are crane and hoist loads, working at height, and handling compressed gas cylinders. This paper outlines the projectile hazard assoicated with compressed gas cylinders and mthods of treatment to provide for compressed gas safety. This information should be of interest to personnel at both magnetic and inertial fusion experiments.

  11. Parametric Thermal Models of the Transient Reactor Test Facility (TREAT)

    Energy Technology Data Exchange (ETDEWEB)

    Bradley K. Heath

    2014-03-01

    This work supports the restart of transient testing in the United States using the Department of Energy’s Transient Reactor Test Facility at the Idaho National Laboratory. It also supports the Global Threat Reduction Initiative by reducing proliferation risk of high enriched uranium fuel. The work involves the creation of a nuclear fuel assembly model using the fuel performance code known as BISON. The model simulates the thermal behavior of a nuclear fuel assembly during steady state and transient operational modes. Additional models of the same geometry but differing material properties are created to perform parametric studies. The results show that fuel and cladding thermal conductivity have the greatest effect on fuel temperature under the steady state operational mode. Fuel density and fuel specific heat have the greatest effect for transient operational model. When considering a new fuel type it is recommended to use materials that decrease the specific heat of the fuel and the thermal conductivity of the fuel’s cladding in order to deal with higher density fuels that accompany the LEU conversion process. Data on the latest operating conditions of TREAT need to be attained in order to validate BISON’s results. BISON’s models for TREAT (material models, boundary convection models) are modest and need additional work to ensure accuracy and confidence in results.

  12. A neutron tomography facility at a low power research reactor

    CERN Document Server

    Körner, S; Von Tobel, P; Rauch, H

    2001-01-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at ...

  13. Incipient Transient Detection in Reactor Systems: Experimental and Theoretical Investigation

    Energy Technology Data Exchange (ETDEWEB)

    Lefteri H. Tsoukalas; S.T. Revankar; X Wang; R. Sattuluri

    2005-09-27

    The main goal of this research was to develop a method for detecting reactor system transients at the earliest possible time through a comprehensive experimental, testing and benchmarking program. This approach holds strong promise for developing new diagnostic technologies that are non-intrusive, generic and highly portable across different systems. It will help in the design of new generation nuclear power reactors, which utilize passive safety systems with a reliable and non-intrusive multiphase flow diagnostic system to monitor the function of the passive safety systems. The main objective of this research was to develop an improved fuzzy logic based detection method based on a comprehensive experimental testing program to detect reactor transients at the earliest possible time, practically at their birth moment. A fuzzy logic and neural network based transient identification methodology and implemented in a computer code called PROTREN was considered in this research and was compared with SPRT (Sequentially Probability Ratio Testing) decision and Bayesian inference. The project involved experiment, theoretical modeling and a thermal-hydraulic code assessment. It involved graduate and undergraduate students participation providing them with exposure and training in advanced reactor concepts and safety systems. In this final report, main tasks performed during the project period are summarized and the selected results are presented. Detailed descriptions for the tasks and the results are presented in previous yearly reports (Revankar et al 2003 and Revankar et al 2004).

  14. Experiment Needs and Facilities Study Appendix A Transient Reactor Test Facility (TREAT) Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    None

    1976-09-01

    The TREAT Upgrade effort is designed to provide significant new capabilities to satisfy experiment requirements associated with key LMFBR Safety Issues. The upgrade consists of reactor-core modifications to supply the physics performance needed for the new experiments, an Advanced TREAT loop with size and thermal-hydraulics capabilities needed for the experiments, associated interface equipment for loop operations and handling, and facility modifications necessary to accommodate operations with the Loop. The costs and schedules of the tasks to be accomplished under the TREAT Upgrade project are summarized. Cost, including contingency, is about 10 million dollars (1976 dollars). A schedule for execution of 36 months has been established to provide the new capabilities in order to provide timely support of the LMFBR national effort. A key requirement for the facility modifications is that the reactor availability will not be interrupted for more than 12 weeks during the upgrade. The Advanced TREAT loop is the prototype for the STF small-bundle package loop. Modified TREAT fuel elements contain segments of graphite-matrix fuel with graded uranium loadings similar to those of STF. In addition, the TREAT upgrade provides for use of STF-like stainless steel-UO{sub 2} TREAT fuel for tests of fully enriched fuel bundles. This report will introduce the Upgrade study by presenting a brief description of the scope, performance capability, safety considerations, cost schedule, and development requirements. This work is followed by a "Design Description". Because greatly upgraded loop performance is central to the upgrade, a description is given of Advanced TREAT loop requirements prior to description of the loop concept. Performance requirements of the upgraded reactor system are given. An extensive discussion of the reactor physics calculations performed for the Upgrade concept study is provided. Adequate physics performance is essential for performance of experiments with

  15. Operating manual for the High Flux Isotope Reactor. Volume I. Description of the facility

    Energy Technology Data Exchange (ETDEWEB)

    1982-09-01

    This volume contains a comprehensive description of the High Flux Isotope Reactor Facility. Its primary purpose is to supplement the detailed operating procedures, providing the reactor operators with background information on the various HFIR systems. The detailed operating procdures are presented in another report.

  16. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

  17. Operation of N Reactor and Fuels Fabrication Facilities, Hanford Reservation, Richland, Benton County, Washington: Environmental assessment

    Energy Technology Data Exchange (ETDEWEB)

    1980-08-01

    Environmental data, calculations and analyses show no significant adverse radiological or nonradiological impacts from current or projected future operations resulting from N Reactor, Fuels Fabrication and Spent Fuel Storage Facilities. Nonoccupational radiation exposures resulting from 1978 N Reactor operations are summarized and compared to allowable exposure limits.

  18. The Jules Horowitz reactor (JHR), a European material testing reactor (MTR), with extended experimental capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A.; Bergamaschi, Y.; Bouilloux, Y.; Bravo, X.; Guigon, B.; Rommens, M.; Tremodeux, P. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, 13 - Saint-Paul-lez-Durance (France)]|[CEA Saclay Dir. de l' Energie Nucleaire DEN, 91 - Gif sur Yvette (France)

    2003-07-01

    The Jules Horowitz Reactor (JHR) is the European MTR (Material Testing Reactor) designed to provide, after 2010, the necessary knowledge for keeping the existing power plants in operation and to design innovative reactors types with new objectives such as: minimizing the radioactive waste production, taking into account additional safety requirements, preventing risks of nuclear proliferation... To achieve such an ambitious objective. The JHR is designed with a high flexibility in order to satisfy the current demand from European industry, research and to be able to accommodate future requirements. The JHR will offer a wide range of performances and services in gathering, in a single site at Cadarache, all the necessary functionalities and facilities for an effective production of results: e.g. fuel fabrication laboratories, preparation of the instrumented devices, interpretation of the experiments, modelling. The JHR must rely on a top level scientific environment based on experts teams from CEA and EC and local universities. With a thermal flux of 7,4.10{sup 14} ncm{sup -2} s{sup -1} and a fast flux of 6,4.10{sup 14} ncm{sup -2}s{sup -1}, it is possible to carry out irradiation experiments on materials and fuels whatever the reactor type considered. It will also be possible to carry out locally, fast neutron irradiation to achieve damage effect up to 25 dpa/year. (dpa = displacement per atom.) The study of the fuels behavior under accidental conditions, from analytical experiments, on a limited amount of irradiated fuel, is a major objective of the project. These oriented safety tests are possible by taking into account specific requirements in the design of the facility such as the tightness level of the containment building, the addition of an alpha hot cell and a laboratory for on line fission products measurement. (authors)

  19. Development of a CELSS Experimental Facility

    Science.gov (United States)

    Guo, S.; Tang, Y.; Zhu, J.; Wang, X.; Yin, Y.; Feng, H.; Ai, W.; Liu, X.; Qin, L.

    A CELSS Experimental Facility was developed two years ago. It contains a volume of about 40.0 m3 and a cultivating area of about 8.4 m2; its interior atmospheric parameters such as temperature, relative humidity, oxygen concentration, carbon dioxide concentration, total pressure, lighting intensity, photoperiod, water content in the growing-matrix, CO2-added accumulative amount, O2-released accumulative amount and ethylene concentration are all controlled and logged automatically and effectively; its growing system consists of two rows of racks along its left-and-right sides separately, each side holds two upper-and-lower layers, and the vertical distance of each growing bed can be adjusted automatically and independently; lighting sources consist of both red (95%) and blue (5%) light-emitting diodes (LED), and the average lighting intensity of each lamp bank at 20-cm distance position under it, reaches to 255.0 μmol m-2 s-1. After that, demonstrating tests were carried out and were finally followed by growing lettuce in the facility. The results showed that all subsystems operated well and all parameters were controlled automatically and efficiently. The lettuce plants in the system could grow much well. Successful development of this system laid a necessary foundation for future larger-scale studies on CELSS integration technique.

  20. Experimental investigation of a new method for advanced fast reactor shutdown cooling

    Science.gov (United States)

    Pakholkov, V. V.; Kandaurov, A. A.; Potseluev, A. I.; Rogozhkin, S. A.; Sergeev, D. A.; Troitskaya, Yu. I.; Shepelev, S. F.

    2017-07-01

    We consider a new method for fast reactor shutdown cooling using a decay heat removal system (DHRS) with a check valve. In this method, a coolant from the decay heat exchanger (DHX) immersed into the reactor upper plenum is supplied to the high-pressure plenum and, then, inside the fuel subassemblies (SAs). A check valve installed at the DHX outlet opens by the force of gravity after primary pumps (PP-1) are shut down. Experimental studies of the new and alternative methods of shutdown cooling were performed at the TISEY test facility at OKBM. The velocity fields in the upper plenum of the reactor model were obtained using the optical particle image velocimetry developed at the Institute of Applied Physics (Russian Academy of Sciences). The study considers the process of development of natural circulation in the reactor and the DHRS models and the corresponding evolution of the temperature and velocity fields. A considerable influence of the valve position in the displacer of the primary pump on the natural circulation of water in the reactor through the DHX was discovered (in some modes, circulation reversal through the DHX was obtained). Alternative DHRS designs without a shell at the DHX outlet with open and closed check valve are also studied. For an open check valve, in spite of the absence of a shell, part of the flow is supplied through the DHX pipeline and then inside the SA simulators. When simulating power modes of the reactor operation, temperature stratification of the liquid was observed, which increased in the cooling mode via the DHRS. These data qualitatively agree with the results of tests at BN-600 and BN-800 reactors.

  1. Experimental Investigation of Effect on Hydrate Formation in Spray Reactor

    Directory of Open Access Journals (Sweden)

    Jianzhong Zhao

    2015-01-01

    Full Text Available The effects of reaction condition on hydrate formation were conducted in spray reactor. The temperature, pressure, and gas volume of reaction on hydrate formation were measured in pure water and SDS solutions at different temperature and pressure with a high-pressure experimental rig for hydrate formation. The experimental data and result reveal that additives could improve the hydrate formation rate and gas storage capacity. Temperature and pressure can restrict the hydrate formation. Lower temperature and higher pressure can promote hydrate formation, but they can increase production cost. So these factors should be considered synthetically. The investigation will promote the advance of gas storage technology in hydrates.

  2. 10 CFR 50.83 - Release of part of a power reactor facility or site for unrestricted use.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Release of part of a power reactor facility or site for... of a power reactor facility or site for unrestricted use. (a) Prior written NRC approval is required... release. Nuclear power reactor licensees seeking NRC approval shall— (1) Evaluate the effect of releasing...

  3. Radiation facilities for fusion-reactor first-wall and blanket structural-materials development

    Energy Technology Data Exchange (ETDEWEB)

    Klueh, R.L.; Bloom, E.E.

    1981-12-01

    Present and future irradiation facilities for the study of fusion reactor irradiation damage are reviewed. Present studies are centered on irradiation in accelerator-based neutron sources, fast- and mixed-spectrum fission reactors, and ion accelerators. The accelerator-based neutron sources are used to demonstrate damage equivalence between high-energy neutrons and fission reactor neutrons. Once equivalence is demonstrated, the large volume of test space available in fission reactors can be used to study displacement damage, and in some instances, the effects of high-helium concentrations and the interaction of displacement damage and helium on properties. Ion bombardment can be used to study the mechanisms of damage evolution and the interaction of displacement damage and helium. These techniques are reviewed, and typical results obtained from such studies are examined. Finally, future techniques and facilities for developing damage levels that more closely approach those expected in an operating fusion reactor are discussed.

  4. Declassification of radioactive water from a pool type reactor after nuclear facility dismantling

    Science.gov (United States)

    Arnal, J. M.; Sancho, M.; García-Fayos, B.; Verdú, G.; Serrano, C.; Ruiz-Martínez, J. T.

    2017-09-01

    This work is aimed to the treatment of the radioactive water from a dismantled nuclear facility with an experimental pool type reactor. The main objective of the treatment is to declassify the maximum volume of water and thus decrease the volume of radioactive liquid waste to be managed. In a preliminary stage, simulation of treatment by the combination of reverse osmosis (RO) and evaporation have been performed. Predicted results showed that the combination of membrane and evaporation technologies would result in a volume reduction factor higher than 600. The estimated time to complete the treatment was around 650 h (25-30 days). For different economical and organizational reasons which are explained in this paper, the final treatment of the real waste had to be reduced and only evaporation was applied. The volume reduction factor achieved in the real treatment was around 170, and the time spent for treatment was 194 days.

  5. Design, construction, and demonstration of a neutron beamline and a neutron imaging facility at a Mark-I TRIGA reactor

    Science.gov (United States)

    Craft, Aaron E.

    The fleet of research and training reactors is aging, and no new research reactors are planned in the United States. Thus, there is a need to expand the capabilities of existing reactors to meet users' needs. While many research reactors have beam port facilities, the original design of the United States Geological Survey TRIGA Reactor (GSTR) did not include beam ports. The MInes NEutron Radiography (MINER) facility developed by this thesis and installed at the GSTR provides new capabilities for both researchers and students at the Colorado School of Mines. The facility consists of a number of components, including a neutron beamline and beamstop, an optical table, an experimental enclosure and associated interlocks, a computer control system, a multi-channel plate imaging detector, and the associated electronics. The neutron beam source location, determined through Monte Carlo modeling, provides the best mixture of high neutron flux, high thermal neutron content, and low gamma radiation content. A Monte Carlo n-Particle (MCNP) model of the neutron beam provides researchers with a tool for designing experiments before placing objects in the neutron beam. Experimental multi-foil activation results, compared to calculated multi-foil activation results, verify the model. The MCNP model predicts a neutron beamline flux of 2.2*106 +/- 6.4*105 n/cm2-s based on a source particle rate determined from the foil activation experiments when the reactor is operating at a power of 950 kWt with the beam shutter fully open. The average cadmium ratio of the beamline is 7.4, and the L/D of the neutron beam is approximately 200+/-10. Radiographs of a sensitivity indicator taken using both the digital detector and the transfer foil method provide one demonstration of the radiographic capabilities of the new facility. Calibration fuel pins manufactured using copper and stainless steel surrogate fuel pellets provide additional specimens for demonstration of the new facility and offer a

  6. STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

    2016-05-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

  7. A Design of Alarm System in a Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jaekwan; Jang, Gwisook; Seo, Sangmun; Suh, Yongsuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The digital alarm system has become an indispensable design to process a large amount of alarms of power plants. Korean research reactor operated for decades maintains a hybrid alarm system with both an analog annunciator and a digital alarm display. In this design, several alarms are indicated on an analog panel and digital display, respectively, and it requires more attention and effort of the operators. As proven in power plants, a centralized alarm system design is necessary for a new research reactor. However, the number of alarms and operators in a research reactor is significantly lesser than power plants. Thus, simplification should be considered as an important factor for the operation efficiency. This paper introduces a simplified alarm system. As advances in information technology, fully digitalized alarm systems have been applied to power plants. In a new research reactor, it will be more useful than an analog or hybrid configuration installed in research reactors decades ago. However, the simplification feature should be considered as an important factor because the number of alarms and number of operators in a research reactor is significantly lesser than in power plants.

  8. Pressure surge in Wendelstein 7-X experimental stellarator facility

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, Algirdas; Uspuras, Eugenijus; Kaliatka, Tadas [Lithuanian Energy Institute, Kaunas (Lithuania). Lab. of Nuclear Installation Safety

    2012-05-15

    Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Wendelstein 7-X (W7-X) is an experimental stellarator facility currently being built in Greifswald, Germany, which shall demonstrate that in the future energy could be produced in such type of fusion reactors. Lithuanian energy institute (LEI) in the frames of European Fusion Development Agreement (EFDA) program is cooperating with Max Planck Institute for Plasma Physics (IPP, Germany) by performing safety analysis of fusion device W7-X. In this paper the consequences of potential water hammer effects were analysed only for the case, when the plasma vessel is operating in the 'baking' mode. 'Baking' is the mode of facility operation during which the plasma vessel structures are heated up to 150 C in order to release absorbed gases from the surfaces and to pump them out of the plasma vessel before plasma operation. For the analysis the thermal-hydraulic model of target (torus) modules cooling / heating systems in W7-X facility was developed using RELAP5/Mod3.3 code. The performed analyses showed that the pressure pulsations in pipelines of cooling / heating systems are possible only in case of very fast closure of automatic valves in torus modules inlets and check valves in torus modules outlets. The maximum dynamic loading to cooling / heating systems pipelines due to such valves activations is in the range 0.12-0.28 MPa. Such dynamic load is insignificant and integrity of pipelines remains not violated. The pressure surge in pipeline connecting torus module in case of erroneous closure of automatic valve is eliminated due to operation of check valve on pipeline in torus module outlet. (orig.)

  9. Pressure surge in Wendelstein 7-X experimental stellarator facility

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, A.; Uspuras, E.; Kaliatka, T. [Lithuanian Energy Inst., Kaunas (Lithuania)

    2011-07-01

    Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Wendelstein 7-X (W7-X) is an experimental stellarator facility currently being built in Greifswald, Germany, which shall demonstrate that in the future energy could be produced in such type of fusion reactors. Lithuanian energy institute (LEI) in the frames of European Fusion Development Agreement (EFDA) program is cooperating with Max Planck Institute for Plasma Physics (IPP, Germany) by performing safety analysis of fusion device W7-X. In this paper the consequences of potential water hammer effects were analysed only for the case, when the plasma vessel is operating in the “baking” mode. “Baking” is the mode of facility operation during which the vacuum vessel structures are heated up to 150 C in order to release absorbed gases from the surfaces and to pump them out of the plasma vessel before plasma operation. For the analysis the thermal-hydraulic model of target (torus) modules cooling / heating systems in W7-X facility was developed using RELAP5/Mod3.3 code. The performed analyses showed that the pressure pulsations in pipelines of cooling / heating systems are possible only in case of very fast closure of automatic valves in torus modules inlets and check valves in torus modules outlets. The maximum dynamic loading to cooling / heating systems pipelines due to such valves activations is in the range 0.12 - 0.28 MPa. Such dynamic load is insignificant and integrity of pipelines remains not violated. The pressure surge in pipeline connecting torus module in case of erroneous closure of automatic valve is eliminated due to operation of check valve on pipeline in torus module outlet. (author)

  10. Generation of an activation map for decommissioning planning of the Berlin Experimental Reactor-II

    Science.gov (United States)

    Lapins, Janis; Guilliard, Nicole; Bernnat, Wolfgang

    2017-09-01

    The BER-II is an experimental facility with 10 MW that was operated since 1974. Its planned operation will end in 2019. To support the decommissioning planning, a map with the overall distribution of relevant radionuclides has to be created according to the state of the art. In this paper, a procedure to create these 3-d maps using a combination of MCNP and deterministic methods is presented. With this approach, an activation analysis is performed for the whole reactor geometry including the most remote parts of the concrete shielding.

  11. Scaling Studies for High Temperature Test Facility and Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schult; Paul D. Bayless; Richard W. Johnson; James R. Wolf; Brian Woods

    2012-02-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5-year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant (NGNP) project. Because the NRC's interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC).

  12. Enclosed Small and Medium Caliber Firing Experimental Facility

    Data.gov (United States)

    Federal Laboratory Consortium — This facility conducts completely instrumented terminal ballistics experimental tests with small and medium-caliber tungsten alloy penetrators against advanced armor...

  13. Experimental Investigation of Biogas Reforming in Gliding Arc Plasma Reactors

    Directory of Open Access Journals (Sweden)

    P. Thanompongchart

    2014-01-01

    Full Text Available Biogas is an important renewable energy source. Its utilization is restricted to vicinity of farm areas, unless pipeline networks or compression facilities are established. Alternatively, biogas may be upgraded into synthetic gas via reforming reaction. In this work, plasma assisted reforming of biogas was investigated. A laboratory gliding arc plasma setup was developed. Effects of CH4/CO2 ratio (1, 2.33, 9, feed flow rate (16.67–83.33 cm3/s, power input (100–600 W, number of reactor, and air addition (0–60% v/v on process performances in terms of yield, selectivity, conversion, and energy consumption were investigated. High power inputs and long reaction time from low flow rates, or use of two cascade reactors were found to promote dry reforming of biogas. High H2 and CO yields can be obtained at low energy consumption. Presence of air enabled partial oxidation reforming that produced higher CH4 conversion, compared to purely dry CO2 reforming process.

  14. Experimental geothermal research facilities study (Phase 0). Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1974-01-01

    The study comprises Phase 0 of a project for Experimental Geothermal Research Facilities. The study focuses on identification of a representative liquid-dominated geothermal reservoir of moderate temperature and salinity, preliminary engineering design of an appropriate energy conversion system, identification of critical technology, and planning for implementation of experimental facilities. The objectives included development of liaison with the industrial sector, to ensure responsiveness to their views in facility requirements and planning, and incorporation of environmental and socioeconomic factors. This Phase 0 report covers problem definition and systems requirements. Facilities will incorporate capability for research in component, system, and materials technology and a nominal 10 MWe experimental, binary cycle, power generating plant.

  15. Advanced Test Reactor National Scientific User Facility 2010 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Mary Catherine Thelen; Todd R. Allen

    2011-05-01

    This is the 2010 ATR National Scientific User Facility Annual Report. This report provides an overview of the program for 2010, along with individual project reports from each of the university principal investigators. The report also describes the capabilities offered to university researchers here at INL and at the ATR NSUF partner facilities.

  16. Investigation of TASS/SMR Capability to Predict a Natural Circulation in the Test Facility for an Integral Reactor

    Directory of Open Access Journals (Sweden)

    Young-Jong Chung

    2014-01-01

    Full Text Available System-integrated modular advanced reactor (SMART is a small-sized advanced integral type pressurized water reactor (PWR with a rated thermal power of 330 MW. It can produce 100 MW of electricity or 90 MW of electricity and 40,000 ton of desalinated water concurrently, which is sufficient for 100,000 residents. The design features contributing to safety enhancement are basically inherent safety improvement and passive safety features. TASS/SMR code was developed for an analysis of design based events and accidents in an integral type reactor reflecting the characteristics of the SMART design. The main purpose of the code is to analyze all relevant phenomena and processes. The code should be validated using experimental data in order to confirm prediction capability. TASS/SMR predicts well the overall thermal-hydraulic behavior under various natural circulation conditions at the experimental test facility for an integral reactor. A pressure loss should be provided a function of Reynolds number at low velocity conditions in order to simulate the mass flow rate well under natural circulations.

  17. Methanol synthesis in a countercurrent gas-solid-solid trickle flow reactor. An experimental study

    NARCIS (Netherlands)

    Kuczynski, M.; Oyevaar, M.H.; Pieters, R.T.; Westerterp, K.R.

    1987-01-01

    The synthesis of methanol from CO and H2 was executed in a gas-solid-solid trickle flow reactor. The reactor consisted of three tubular reactor sections with cooling sections in between. The catalyst was Cu on alumina, the adsorbent was a silica-alumina powder and the experimental range 498–523 K,

  18. Reed Reactor Facility final report, September 1, 1995--August 31, 1996

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1995 to August 31, 1996. This report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission, the US Department of Energy, and the Oregon Department of Energy. Highlights of the last year include: student participation in the program is very high; the facility continues its success in obtaining donated equipment from the Portland General Electric, US Department of Energy, and other sources; the facility is developing more paid work; progress is being made in a collaborative project with Pacific Northwest National Laboratory on isotope production for medical purposes. There were over 1,500 individual visits to the Reactor Facility during the year. Most were students in classes at Reed College or area universities, colleges, and high schools. Including tours and research conducted at the facility, the Reed Reactor Facility contributed to the educational programs of six colleges and universities in addition to eighteen pre-college groups. During the year, the reactor was operated almost three hundred separate times. The total energy production was over 23 MW-hours. The reactor staff consists of a Director, an Associated Director, a contract Health Physicist, and approximately twenty Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 5% of the federal limits.

  19. Experimental and numerical stability investigations on natural circulation boiling water reactors

    NARCIS (Netherlands)

    Marcel, C.P.

    2007-01-01

    The stability of natural circulation boiling water reactors is investigated with a strong emphasis on experiments. Two different facilities are used for such a task: the GENESIS facility (to which a void reactivity feedback system is artificially added) and the CIRCUS facility. In addition,

  20. Upgrading of neutron radiography/tomography facility at research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abd El Bar, Waleed; Mongy, Tarek [Atomic Energy Authority, Cairo (Egypt). ETRR-2; Kardjilov, Nikolay [Helmholtz Zentrum Berlin (HZB) for Materials and Energy, Berlin (Germany)

    2014-03-15

    A state-of-the-art neutron tomography imaging system was set up at the neutron radiography beam tube at the Egypt Second Research Reactor (ETRR-2) and was successfully commissioned in 2013. This study presents a set of tomographic experiments that demonstrate a high quality tomographic image formation. A computer technique for data processing and 3D image reconstruction was used to see inside a copy module of an ancient clay article provided by the International Atomic Energy Agency (IAEA). The technique was also able to uncover tomographic imaging details of a mummified fish and provided a high resolution tomographic image of a defective fire valve. (orig.)

  1. TREAT (Transient Reactor Test Facility) reactor control rod scram system simulations and testing

    Energy Technology Data Exchange (ETDEWEB)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs.

  2. Experimental Evaluation of a Water Shield for a Surface Power Reactor

    Science.gov (United States)

    Pearson, J. Boise; Reid, Robert S.

    2006-01-01

    As part of the Vision for Space Exploration the end of the next decade will bring man back to the surface of the moon. One of the most critical issues for the establishment of human presence on the moon will be the availability of compact power sources. The establishment of man on the moon will require power from greater than 10's of kWt's in follow on years. Nuclear reactors are extremely we11 suited to meet the needs for power generation on the lunar or Martian surface. reactor system. Several competing concepts exist for lightweight, safe, robust shielding systems such as a water shield, lithium hydride (LiH), Boron Carbide, and others. Water offers several potential advantages, including reduced cost, reduced technical risk, and reduced mass. Water has not typically been considered for space reactor applications because of the need for gravity to remove the potential for radiation streaming paths. The water shield concept relies on predictions of passive circulation of the shield water by natural convection to adequately cool the shield. This prediction needs to be experimentally evaluated, especially for shields with complex geometries. MSFC has developed the experience and fac necessary to do this evaluation in the Early Flight Fission - Test Facility (EFF-TF).

  3. Overview of the Neutron experimental facilities at LANSCE

    Energy Technology Data Exchange (ETDEWEB)

    Mocko, Michal [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-06-30

    This presentation gives an overview of the neutron experimental facilities at LANSCE. The layout is mentioned in detail, with a map of the south-side experimental facilities, information on Target-4 and the Lujan Center. Then it goes into detail about neutron sources, specifically continuous versus pulsed. Target 4 is then discussed. In conclusion, we have introduced the south-side experimental facilities in operation at LANSCE. 1L target and Target 4 provide complementary neutron energy spectra. Two spallation neutron sources taken together cover more than 11 orders of magnitude in neutron energy.

  4. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

    2015-04-29

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  5. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Science.gov (United States)

    Sulaiman, S. A.; Dominguez-Ontiveros, E. E.; Alhashimi, T.; Budd, J. L.; Matos, M. D.; Hassan, Y. A.

    2015-04-01

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A&M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  6. A comparison of different neutron spectroscopy systems at the reactor facility VENUS

    CERN Document Server

    Vanhavere, F; Chartier, J L; Itie, C; Rosenstock, W; Koeble, T; D'Errico, F

    2002-01-01

    The VENUS facility is a zero-power research reactor mainly devoted to studies on LWR fuels. Localised high-neutron rates were found around the reactor, with a neutron/gamma dose equivalent rate ratio as high as three. Therefore, a study of the neutron dosimetry around the reactor was started some years ago. During this study, several methods of neutron spectroscopy were employed and a study of individual and ambient dosemeters was performed. A first spectrometric measurement was done with the IPSN multisphere spectrometer in three positions around the reactor. Secondly, the ROSPEC spectrometer from the Fraunhofer Institut was used. The spectra were also measured with the bubble interactive neutron spectrometer. These measurements were compared with a numerical simulation of the neutron field made with the code TRIPOLI-3. Dosimetric measurements were made with three types of personal neutron dosemeters: an albedo type, a track etch detector and a bubble detector.

  7. Preliminary Analysis of the Transient Reactor Test Facility (TREAT) with PROTEUS

    Energy Technology Data Exchange (ETDEWEB)

    Connaway, H. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-11-30

    The neutron transport code PROTEUS has been used to perform preliminary simulations of the Transient Reactor Test Facility (TREAT). TREAT is an experimental reactor designed for the testing of nuclear fuels and other materials under transient conditions. It operated from 1959 to 1994, when it was placed on non-operational standby. The restart of TREAT to support the U.S. Department of Energy’s resumption of transient testing is currently underway. Both single assembly and assembly-homogenized full core models have been evaluated. Simulations were performed using a historic set of WIMS-ANL-generated cross-sections as well as a new set of Serpent-generated cross-sections. To support this work, further analyses were also performed using additional codes in order to investigate particular aspects of TREAT modeling. DIF3D and the Monte-Carlo codes MCNP and Serpent were utilized in these studies. MCNP and Serpent were used to evaluate the effect of geometry homogenization on the simulation results and to support code-to-code comparisons. New meshes for the PROTEUS simulations were created using the CUBIT toolkit, with additional meshes generated via conversion of selected DIF3D models to support code-to-code verifications. All current analyses have focused on code-to-code verifications, with additional verification and validation studies planned. The analysis of TREAT with PROTEUS-SN is an ongoing project. This report documents the studies that have been performed thus far, and highlights key challenges to address in future work.

  8. Experimental and MCNP5 based evaluation of neutron and gamma flux in the irradiation ports of the University of Utah research reactor

    Directory of Open Access Journals (Sweden)

    Noble Brooklyn

    2012-01-01

    Full Text Available Neutron and gamma flux environment of various irradiation ports in the University of Utah training, research, isotope production, general atomics reactor were experimentally assessed and fully modeled using the MCNP5 code. The experimental measurements were based on the cadmium ratio in the irradiation ports of the reactor, flux profiling using nickel wire, and gamma dose measurements using thermo luminescence dosimeter. Full 3-D MCNP5 reactor model was developed to obtain the neutron flux distributions of the entire reactor core and to compare it with the measured flux focusing at the irradiation ports. Integration of all these analysis provided the updated comprehensive neutron-gamma flux maps of the existing irradiation facilities of the University of Utah TRIGA reactor.

  9. Operational Philosophy for the Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    J. Benson; J. Cole; J. Jackson; F. Marshall; D. Ogden; J. Rempe; M. C. Thelen

    2013-02-01

    In 2007, the Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF). At its core, the ATR NSUF Program combines access to a portion of the available ATR radiation capability, the associated required examination and analysis facilities at the Idaho National Laboratory (INL), and INL staff expertise with novel ideas provided by external contributors (universities, laboratories, and industry). These collaborations define the cutting edge of nuclear technology research in high-temperature and radiation environments, contribute to improved industry performance of current and future light-water reactors (LWRs), and stimulate cooperative research between user groups conducting basic and applied research. To make possible the broadest access to key national capability, the ATR NSUF formed a partnership program that also makes available access to critical facilities outside of the INL. Finally, the ATR NSUF has established a sample library that allows access to pre-irradiated samples as needed by national research teams.

  10. Approaches to experimental validation of high-temperature gas-cooled reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Belov, S.E. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Borovkov, M.N., E-mail: borovkov@okbm.nnov.ru [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Golovko, V.F.; Dmitrieva, I.V.; Drumov, I.V.; Znamensky, D.S.; Kodochigov, N.G. [Joint Stock Company ' Afrikantov OKB Mechanical Engineering' , Burnakovsky Proezd, 15, Nizhny Novgorod 603074 (Russian Federation); Baxi, C.B.; Shenoy, A.; Telengator, A. [General Atomics, 3550 General Atomics Court, CA (United States); Razvi, J., E-mail: Junaid.Razvi@ga.com [General Atomics, 3550 General Atomics Court, CA (United States)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Computational and experimental investigations of thermal and hydrodynamic characteristics for the equipment. Black-Right-Pointing-Pointer Vibroacoustic investigations. Black-Right-Pointing-Pointer Studies of the electromagnetic suspension system on GT-MHR turbo machine rotor models. Black-Right-Pointing-Pointer Experimental investigations of the catcher bearings design. - Abstract: The special feature of high-temperature gas-cooled reactors (HTGRs) is stressed operating conditions for equipment due to high temperature of the primary circuit helium, up to 950 Degree-Sign C, as well as acoustic and hydrodynamic loads upon the gas path elements. Therefore, great significance is given to reproduction of real operation conditions in tests. Experimental investigation of full-size nuclear power plant (NPP) primary circuit components is not practically feasible because costly test facilities will have to be developed for the power of up to hundreds of megawatts. Under such conditions, the only possible process to validate designs under development is representative tests of smaller scale models and fragmentary models. At the same time, in order to take in to validated account the effect of various physical factors, it is necessary to ensure reproduction of both individual processes and integrated tests incorporating needed integrated investigations. Presented are approaches to experimental validation of thermohydraulic and vibroacoustic characteristics for main equipment components and primary circuit path elements under standard loading conditions, which take account of their operation in the HTGR. Within the framework of the of modular helium reactor project, including a turbo machine in the primary circuit, a new and difficult problem is creation of multiple-bearing flexible vertical rotor. Presented are approaches to analytical and experimental validation of the rotor electromagnetic bearings, catcher bearings, flexible rotor

  11. Testing and Research Capabilities at the Sandia Fast Pulsed Reactor Facility

    Science.gov (United States)

    Berry, Donald T.

    1994-07-01

    A wide variety of space-based system components have been qualified for use through neutron irradiation testing performed at the Sandia Pulsed Reactor (SPR) Facility. The SPR Facility is the operating location for two fast burst reactors, SPR II and SPR III, which have been used to induce neutron and gamma damage in electronic components and other materials for customers in the Department of Energy, Department of Defense, NASA, and the private sector. In addition to the pulse mode of operation, during which peak fluxes of up to 1023 n/m2-s are achieved, the steady state mode allows for the long term irradiation of components and systems in a fast neutron environment at a flux of up to 5×1015 n/m2-s. The SPR reactors are operated in a 9.2 meter diameter exposure cell, or Kiva, suitable for the irradiation of large test articles external to the reactors. Currently, a new upgraded version of SPR III (SPR HIM) is in fabrication; a unique feature of SPR HIM is its 190 mm (usable diameter) central irradiation cavity, the largest of any U.S. fast burst reactor. An improved cooling system permits continuous operation at power levels in excess of 20 kWt. The SPR Facility is also the operating site for a critical assembly which was used to characterize prototypic fuels in arrays appropriate for the Space Nuclear Thermal Propulsion Program. Work continues on use of the facility to design, build, and operate critical assemblies for a diverse customer base.

  12. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  13. Design and R&D Progress of China Lead-Based Reactor for ADS Research Facility

    Directory of Open Access Journals (Sweden)

    Yican Wu

    2016-03-01

    Full Text Available In 2011, the Chinese Academy of Sciences launched an engineering project to develop an accelerator-driven subcritical system (ADS for nuclear waste transmutation. The China Lead-based Reactor (CLEAR, proposed by the Institute of Nuclear Energy Safety Technology, was selected as the reference reactor for ADS development, as well as for the technology development of the Generation IV lead-cooled fast reactor. The conceptual design of CLEAR-I with 10 MW thermal power has been completed. KYLIN series lead-bismuth eutectic experimental loops have been constructed to investigate the technologies of the coolant, key components, structural materials, fuel assembly, operation, and control. In order to validate and test the key components and integrated operating technology of the lead-based reactor, the lead alloy-cooled non-nuclear reactor CLEAR-S, the lead-based zero-power nuclear reactor CLEAR-0, and the lead-based virtual reactor CLEAR-V are under realization.

  14. Tokamak experimental power reactor conceptual design. Volume I

    Energy Technology Data Exchange (ETDEWEB)

    1976-08-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 years. The EPR operates in a pulsed mode at a frequency of approximately 1/min., with an approximate 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2-cm thick stainless steel, and has 2-cm thick detachable, beryllium-coated coolant panels mounted on the interior. An 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H/sub 2/O. Sixteen niobium-titanium superconducting toroidal-field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic-heating and equilibrium-field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam-injectors, which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-converters.

  15. Reed Reactor Facility final report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-09-01

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: student participation in the program is very high; the facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources; the facility is developing more paid work. There were 1,115 visits of the Reactor Facility by individuals during the year. Most of these visitors were students in classes at Reed College or area universities, colleges, and high schools. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Associate Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below 1% of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits.

  16. CFD and experimental investigation of sloshing parameters for the safety assessment of HLM reactors

    Energy Technology Data Exchange (ETDEWEB)

    Myrillas, Konstantinos, E-mail: myrillas@vki.ac.be [von Karman Institute for Fluid Dynamics, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Planquart, Philippe, E-mail: philippe.planquart@vki.ac.be [von Karman Institute for Fluid Dynamics, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Simonini, Alessia, E-mail: Simonini@vki.ac.be [von Karman Institute for Fluid Dynamics, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Buchlin, Jean-Marie, E-mail: buchlin@vki.ac.be [von Karman Institute for Fluid Dynamics, Chaussée de Waterloo 72, B-1640 Rhode-St-Genèse (Belgium); Schyns, Marc, E-mail: mschyns@SCKCEN.BE [SCK-CEN, Boeretang 200, B-2400 Mol (Belgium)

    2017-02-15

    Highlights: • Comparison of sloshing behavior in cylindrical tank using mercury and water. • Flow visualization of liquid sloshing in resonance case. • CFD simulations of sloshing with OpenFOAM, using the VOF method. • Qualitative and quantitative comparison of experimental and numerical results. • Evaluation of sloshing forces on the tank walls from numerical simulations. - Abstract: For the safety assessment of Heavy Liquid Metal nuclear reactors under seismic excitation, sloshing phenomena can be of great concern. The earthquake motions are transferred to the liquid coolant which oscillates inside the vessel, exerting additional forces on the walls and internal structures. The present study examines the case of MYRRHA, a multi-purpose experimental reactor with LBE as coolant, developed by SCK·CEN. The sloshing behavior of liquid metals is studied through a comparison between mercury and water in a cylindrical tank. Experimental investigation of sloshing is carried out using optical techniques with the shaking table facility SHAKESPEARE at the von Karman Institute. Emphasis is given on the resonance case, where maximum forces occur on the tank walls. The experimental cases are reproduced numerically with the CFD software OpenFOAM, using the VOF method to track the liquid interface. The non-linear nature of sloshing is observed through visualization, where swirling is shown in the resonance case. The complex behavior is well reproduced by the CFD simulations, providing good qualitative validation of the numerical tools. A quantitative comparison of the maximum liquid elevation inside the tank shows higher values for the liquid metal than for water. Some discrepancies are revealed in CFD results and the differences are quantified. From simulations it is verified that the forces scale with the density ratio, following similar evolution in time. Overall, water is demonstrated to be a valid option as a working liquid in order to evaluate the sloshing

  17. Evaluation of integral continuing experimental capability (CEC) concepts for light water reactor research: PWR scaling concepts

    Energy Technology Data Exchange (ETDEWEB)

    Condie, K G; Larson, T K; Davis, C B; McCreery, G E

    1987-02-01

    In this report reactor transients and thermal-hydraulic phenomena of importance (based on probabilistic risk assessment and the International Code Assessment Program) to reactor safety were examined and identified. Established scaling methodologies were used to develop potential concepts for integral thermal-hydraulic testing facilities. Advantages and disadvantages of each concept are evaluated. Analysis is conducted to examine the scaling of various phenomena in each of the selected concepts. Results generally suggest that a facility capable of operating at typical reactor operating conditions will scale most phenomena reasonably well. Although many phenomena in facilities using Freon or water at nontypical pressure will scale reasonably well, those phenomena that are heavily dependent on quality (heat transfer or critical flow for example) can be distorted. Furthermore, relation of data produced in facilities operating with nontypical fluids or at nontypical pressures to large plants will be a difficult and time consuming process.

  18. Development of a fault test experimental facility model using Matlab

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, Iraci Martinez; Moraes, Davi Almeida, E-mail: martinez@ipen.br, E-mail: dmoraes@dk8.com.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The Fault Test Experimental Facility was developed to simulate a PWR nuclear power plant and is instrumented with temperature, level and pressure sensors. The Fault Test Experimental Facility can be operated to generate normal and fault data, and these failures can be added initially small, and their magnitude being increasing gradually. This work presents the Fault Test Experimental Facility model developed using the Matlab GUIDE (Graphical User Interface Development Environment) toolbox that consists of a set of functions designed to create interfaces in an easy and fast way. The system model is based on the mass and energy inventory balance equations. Physical as well as operational aspects are taken into consideration. The interface layout looks like a process flowchart and the user can set the input variables. Besides the normal operation conditions, there is the possibility to choose a faulty variable from a list. The program also allows the user to set the noise level for the input variables. Using the model, data were generated for different operational conditions, both under normal and fault conditions with different noise levels added to the input variables. Data generated by the model will be compared with Fault Test Experimental Facility data. The Fault Test Experimental Facility theoretical model results will be used for the development of a Monitoring and Fault Detection System. (author)

  19. Introduction of Nuclear Instrumentations and Radiation Measurements in Experimental Fast Reactor 「JOYO」

    OpenAIRE

    大戸 敏弘; 鈴木 惣十

    1992-01-01

    This report introduces the nuclear instrumentation system and major R&D (research and development) activities using radiation measurement techniques in Experimental Fast Reactor "JOYO". In the introduction of the nuclear instrumentation system, following items are described; (1)system function (2)roles as a reactor plant equipment (3)specifications and charactelistics of neutron detectors, (4)construction and layout of the system. For reactor dosimetry at various irradiation tests and surveil...

  20. Development of Pneumatic Transfer Irradiation Facility (PTS no.3) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-04-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide. The pneumatic transfer irradiation system (PTS no.3) involving a manual system and an semi-automatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor and NAA laboratory of RI building in 2006. In this technical report, the design, operation and control of these system (PTS no.3) was described. Also the experimental results and the characteristic parameters measured from a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  1. Development of Pneumatic Transfer Irradiation Facility (PTS no.1) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer system (PTS no.1) involving a manual system and an semiautomatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of these system (PTS no.1) was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  2. Development of Pneumatic Transfer Irradiation Facility (PTS no.2) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer irradiation system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer irradiation system (PTS no.2) involving a manual system and an automatic system for delayed neutron activation analysis (DNAA) were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of PTS no.2 was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, automatic operation control by personal computer, delayed neutron counting system, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  3. Reed Reactor Facility annual report, September 1, 1994--August 31, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-12-31

    This report covers the period from September 1, 1994 to August 31, 1995. Information contained in this report is intended to fulfill several purposes including the reporting requirements of the US Nuclear Regulatory Commission (USNRC), the US Department of Energy (USDOE), and the Oregon Department of Energy (ODOE). Highlights of the last year include: (1) The number of new licensed student operators more than replaced the number of graduating seniors. Seven Reed College seniors used the reactor as part of their thesis projects. (2) The facility has been extraordinarily successful in obtaining donated equipment from Portland General Electric, US Department of Energy, Precision Castparts, Tektronix, and other sources. Battelle (Pacific Northwest Laboratory) has been generous in lending valuable equipment to the college. (3) The facility is developing more paid work. Income in the past academic year was much greater than the previous year, and next year should increase by even more. Additionally, the US Department of Energy`s Reactor-Use Sharing grant increased significantly this year. During the year, the reactor was operated 225 separate times on 116 days. The total energy production was 24.6 MW-hours. The reactor staff consists of a Director, an Assistant Director, a contract Health Physicist, and approximately fifteen Reed College undergraduate students as hourly employees. All radiation exposures to individuals during this year were well below one percent of the federal limits. There were no releases of liquid radioactive material from the facility and airborne releases (primarily {sup 41}Ar) were well within regulatory limits. No radioactive waste was shipped from the facility during this period.

  4. Guidelines for preparing criticality safety evaluations at Department of Energy non-reactor nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-09-01

    This Department of Energy (DOE) is approved for use by all components of DOE. It contains guidelines that should be followed when preparing Criticality Safety Evaluations that will be used to demonstrate the safety of operations performed at DOE Non-Reactor Nuclear Facilities. Adherence with these guidelines will provide consistency and uniformity in Criticality Safety Evaluations (CSEs) across the complex and will document compliance with DOE Order 5480.24 requirements as they pertain to CSEs.

  5. Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D.; Nuclear Engineering Division

    2005-09-01

    As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: (a) in the RCCS, strong

  6. Review of accident analyses of RB experimental reactor

    Directory of Open Access Journals (Sweden)

    Pešić Milan P.

    2003-01-01

    Full Text Available The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VTNCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62 yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin consisting of 2% enriched uranium metal and 80% enriched UO2 dispersed in aluminum matrix, have been available since 1962 and 1976 respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINĆA Institute, an independent regulatory body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given.

  7. A reverse flow catalytic membrane reactor for the production of syngas: an experimental study

    NARCIS (Netherlands)

    Smit, J.; Bekink, G.J.; van Sint Annaland, M.; Kuipers, J.A.M.

    2005-01-01

    In this paper experimental results are presented for a demonstration unit of a recently proposed novel integrated reactor concept (Smit et. al., 2005) for the partial oxidation of natural gas to syngas (POM), namely a Reverse Flow Catalytic Membrane Reactor (RFCMR). Natural gas has great potential

  8. Experimental and simulated dosimetry of the university of Utah TRIGA reactor

    Science.gov (United States)

    Marble, Benjamin James

    Simulated neutron and gamma transport enable the gamma dose to be estimated at the surface of the University of Utah TRIGA Reactor UUTR pool. These results are benchmarked against experimental results for model verification. This model is useful for future licensing and possible reactor power upgrades. MCNP5 was utilized for the UUTR simulation and comparison with thermoluminescent detectors TLDs.

  9. An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

    1989-12-01

    Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

  10. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    Energy Technology Data Exchange (ETDEWEB)

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  11. COMPUTATIONAL AND EXPERIMENTAL MODELING OF SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Paul C.K. Lam; Isaac K. Gamwo; Dimitri Gidaspow

    2002-05-01

    The objective of this study was to develop a predictive experimentally verified computational fluid dynamics (CFD) model for gas-liquid-solid flow. A three dimensional transient computer code for the coupled Navier-Stokes equations for each phase was developed and is appended in this report. The principal input into the model is the viscosity of the particulate phase which was determined from a measurement of the random kinetic energy of the 800 micron glass beads and a Brookfield viscometer. The details are presented in the attached paper titled ''CFD Simulation of Flow and Turbulence in a Slurry Bubble Column''. This phase of the work is in press in a referred journal (AIChE Journal, 2002) and was presented at the Fourth International Conference on Multiphase Flow (ICMF 2001) in New Orleans, May 27-June 1, 2001 (Paper No. 909). The computed time averaged particle velocities and concentrations agree with Particle Image Velocimetry (PIV) measurements of velocities and concentrations, obtained using a combination of gamma-ray and X-ray densitometers, in a slurry bubble column, operated in the bubbly-coalesced fluidization regime with continuous flow of water. Both the experiment and the simulation show a down-flow of particles in the center of the column and up-flow near the walls and nearly uniform particle concentration. Normal and shear Reynolds stresses were constructed from the computed instantaneous particle velocities. The PIV measurement and the simulation produced instantaneous particle velocities. The PIV measurement and the simulation produced similar nearly flat horizontal profiles of turbulent kinetic energy of particles. To better understand turbulence we studied fluidization in a liquid-solid bed. This work was also presented at the Fourth International Conference on Multiphase Flow (ICMF 2001, Paper No. 910). To understand turbulence in risers, measurements were done in the IIT riser with 530 micron glass beads using a PIV

  12. Post 9-11 Security Issues for Non-Power Reactor Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Zaffuts, P. J.

    2003-02-25

    This paper addresses the legal and practical issues arising out of the design and implementation of a security-enhancement program for non power reactor nuclear facilities. The security enhancements discussed are derived from the commercial nuclear power industry's approach to security. The nuclear power industry's long and successful experience with protecting highly sensitive assets provides a wealth of information and lessons that should be examined by other industries contemplating security improvements, including, but not limited to facilities using or disposing of nuclear materials. This paper describes the nuclear industry's approach to security, the advantages and disadvantages of its constituent elements, and the legal issues that facilities will need to address when adopting some or all of these elements in the absence of statutory or regulatory requirements to do so.

  13. Safety evaluation report related to the renewal of the facility license for the research reactor at the Dow Chemical Company

    Energy Technology Data Exchange (ETDEWEB)

    1989-04-01

    This safety evaluation report for the application filed by the Dow Chemical Company for renewal of facility Operating License R-108 to continue to operate its research reactor at an increased operating power level has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is located on the grounds of the Michigan Division of the Dow Chemical Company in Midland, Michigan. The staff concludes that the Dow Chemical Company can continue to operate its reactor without endangering the health and safety of the public.

  14. Gas cooled fast breeder reactor design for a circulator test facility (modified HTGR circulator test facility)

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    A GCFR helium circulator test facility sized for full design conditions is proposed for meeting the above requirements. The circulator will be mounted in a large vessel containing high pressure helium which will permit testing at the same power, speed, pressure, temperature and flow conditions intended in the demonstration plant. The electric drive motor for the circulator will obtain its power from an electric supply and distribution system in which electric power will be taken from a local utility. The conceptual design decribed in this report is the result of close interaction between the General Atomic Company (GA), designer of the GCFR, and The Ralph M. Parson Company, architect/engineer for the test facility. A realistic estimate of total project cost is presented, together with a schedule for design, procurement, construction, and inspection.

  15. Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

    1994-10-01

    This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

  16. Summary on experimental facilities and future developments at SINQ

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, G.S. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-11-01

    With 13 experimental facilities under construction to become available during the first year of SINQ operation, a nearly complete suite of options for users will be made available to carry out research with neutrons at PSI. Three more facilities are under design and will come on line somewhat later. To complete the suite, three more specialized instruments are being evaluated. SINQ being a novel neutron source concept, significant scope for improvement is also seen on the source side. It is a major goal of PSI to exploit these opportunities and to make - among others - use of neutron instruments to carry out the necessary research. (author) 9 figs., 1 tab., 11 refs.

  17. Hazards assessment for the Waste Experimental Reduction Facility

    Energy Technology Data Exchange (ETDEWEB)

    Calley, M.B.; Jones, J.L. Jr.

    1994-09-19

    This report documents the hazards assessment for the Waste Experimental Reduction Facility (WERF) located at the Idaho National Engineering Laboratory, which is operated by EG&G Idaho, Inc., for the US Department of Energy (DOE). The hazards assessment was performed to ensure that this facility complies with DOE and company requirements pertaining to emergency planning and preparedness for operational emergencies. DOE Order 5500.3A requires that a facility-specific hazards assessment be performed to provide the technical basis for facility emergency planning efforts. This hazards assessment was conducted in accordance with DOE Headquarters and DOE Idaho Operations Office (DOE-ID) guidance to comply with DOE Order 5500.3A. The hazards assessment identifies and analyzes hazards that are significant enough to warrant consideration in a facility`s operational emergency management program. This hazards assessment describes the WERF, the area surrounding WERF, associated buildings and structures at WERF, and the processes performed at WERF. All radiological and nonradiological hazardous materials stored, used, or produced at WERF were identified and screened. Even though the screening process indicated that the hazardous materials could be screened from further analysis because the inventory of radiological and nonradiological hazardous materials were below the screening thresholds specified by DOE and DOE-ID guidance for DOE Order 5500.3A, the nonradiological hazardous materials were analyzed further because it was felt that the nonradiological hazardous material screening thresholds were too high.

  18. 77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

    Science.gov (United States)

    2012-11-15

    ... COMMISSION The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R... Operating License No. R-84 (Application), which currently authorizes the Armed Forces Radiobiology Research... the renewal of Facility Operating License No. R-84, which currently authorizes the licensee to operate...

  19. 77 FR 7613 - Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108

    Science.gov (United States)

    2012-02-13

    ... COMMISSION Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108... renewal of Facility Operating License No. R-108 (``Application''), which currently authorizes the Dow... Operating License No. R-108 for the DTRR. The application contains SUNSI. Based on its initial review of the...

  20. Decommissioning an Active Historical Reactor Facility at the Savannah River Site - 13453

    Energy Technology Data Exchange (ETDEWEB)

    Bergren, Christopher L.; Long, J. Tony; Blankenship, John K. [Savannah River Nuclear Solutions, LLC, Bldg. 730-4B, Aiken, SC 29808 (United States); Adams, Karen M. [United States Department of Energy, Bldg. 730-B, Aiken, SC 29808 (United States)

    2013-07-01

    The Savannah River Site (SRS) is an 802 square-kilometer United States Department of Energy (US DOE) nuclear facility located along the Savannah River near Aiken, South Carolina, where Management and Operations are performed by Savannah River Nuclear Solutions (SRNS). In 2004, DOE recognized SRS as structure within the Cold War Historic District of national, state and local significance composed of the first generation of facilities constructed and operated from 1950 through 1989 to produce plutonium and tritium for our nation's defense. DOE agreed to manage the SRS 105-C Reactor Facility as a potentially historic property due to its significance in supporting the U.S. Cold War Mission and for potential for future interpretation. This reactor has five primary areas within it, including a Disassembly Basin (DB) that received irradiated materials from the reactor, cooled them and prepared the components for loading and transport to a Separation Canyon for processing. The 6,317 square meter area was divided into numerous work/storage areas. The walls between the individual basin compartments have narrow vertical openings called 'slots' that permit the transfer of material from one section to another. Data indicated there was over 830 curies of radioactivity associated with the basin sediments and approximately 9.1 M liters of contaminated water, not including a large quantity of activated reactor equipment, scrap metal, and debris on the basin floor. The need for an action was identified in 2010 to reduce risks to personnel in the facility and to eliminate the possible release of contaminants into the environment. The release of DB water could potentially migrate to the aquifer and contaminate groundwater. DOE, its regulators [U. S. Environmental Protection Agency (USEPA)-Region 4 and the South Carolina Department of Health and Environmental Control (SCDHEC)] and the SC Historical Preservation Office (SHPO) agreed/concurred to perform a non

  1. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  2. Passive and Active Radiation Measurements Capability at the INL Zero Power Physics Reactor (ZPPR) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Robert Neibert; John Zabriskie; Collin Knight; James L. Jones

    2010-12-01

    The Zero Power Physics Reactor (ZPPR) facility is a Department of Energy facility located in the Idaho National Laboratory’s (INL) Materials and Fuels Complex. It contains various nuclear and non-nuclear materials that are available to support many radiation measurement assessments. User-selected, single material, nuclear and non-nuclear materials can be readily utilized with ZPPR clamshell containers with almost no criticality concerns. If custom, multi-material configurations are desired, the ZPPR clamshell or an approved aluminum Inspection Object (IO) Box container may be utilized, yet each specific material configuration will require a criticality assessment. As an example of the specialized material configurations possible, the National Nuclear Security Agency’s Office of Nuclear Verification (NNSA/NA 243) has sponsored the assembly of six material configurations. These are shown in the Appendixes and have been designated for semi-permanent storage that can be available to support various radiation measurement applications.

  3. Presentation and comparison of experimental critical heat flux data at conditions prototypical of light water small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Greenwood, M.S., E-mail: 1greenwoodms@ornl.gov; Duarte, J.P.; Corradini, M.

    2017-06-15

    Highlights: • Low mass flux and moderate to high pressure CHF experimental results are presented. • Facility uses chopped-cosine heater profile in a 2 × 2 square bundle geometry. • The EPRI, CISE-GE, and W-3 CHF correlations provide reasonable average CHF prediction. • Neural network analysis predicts experimental data and demonstrates utility of method. - Abstract: The critical heat flux (CHF) is a two-phase flow phenomenon which rapidly decreases the efficiency of the heat transfer performance at a heated surface. This phenomenon is one of the limiting criteria in the design and operation of light water reactors. Deviations of operating parameters greatly alters the CHF condition and must be experimentally determined for any new parameters such as those proposed in small modular reactors (SMR) (e.g. moderate to high pressure and low mass fluxes). Current open literature provides too little data for functional use at the proposed conditions of prototypical SMRs. This paper presents a brief summary of CHF data acquired from an experimental facility at the University of Wisconsin-Madison designed and built to study CHF at high pressure and low mass flux ranges in a 2 × 2 chopped cosine rod bundle prototypical of conceptual SMR designs. The experimental CHF test inlet conditions range from pressures of 8–16 MPa, mass fluxes of 500–1600 kg/m2 s, and inlet water subcooling from 250 to 650 kJ/kg. The experimental data is also compared against several accepted prediction methods whose application ranges are most similar to the test conditions.

  4. ESTIMATION OF ROUTINE DISCHARGE OF RADIONUCLIDES ON POWER REACTOR EXPERIMENTAL RDE

    OpenAIRE

    Udiyani, P.M; Sri Kuntjoro

    2017-01-01

    Experimental power reactor (RDE) which is planned to be constructed by BATAN is a kind of High Temperature Gas Cooled Reactor (HTGR) with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products remained in the core. Although the fission products released into the environment are very small, in order to comply the regulations the study about environmental radiation on normal or routine operation condition need to be performed. Estimati...

  5. Experimental facility for analysis of biomass combustion characteristics

    Directory of Open Access Journals (Sweden)

    Miljković Biljana M.

    2015-01-01

    Full Text Available The objective of the present article is to present an experimental facility which was designed and built at the Faculty of Technical Sciences in order to study the combustion of different sorts of biomass and municipal solid waste. Despite its apparent simplicity, direct combustion is a complex process from a technological point of view. Conventional combustion equipment is not designed for burning agricultural residues. Devices for agricultural waste combustion are still in the development phase, which means that adequate design solution is presently not available at the world market. In order to construct a boiler and achieve optimal combustion conditions, it is necessary to develop a mathematical model for biomass combustion. Experimental facility can be used for the collection of data necessary for detailed modelling of real grate combustor of solid biomass fuels. Due to the complexity of the grate combustion process, its mathematical models and simulation software tools must be developed and verified using experimental data. This work highlights the properties required for the laboratory facility designed for the examination of biomass combustion and discusses design and operational issues.

  6. Experimental Study of the Effect of Graphite Dispersion on the Heat Transfer Phenomena in a Reactor Cavity Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Vaghetto, Rodolfo; Capone, Luigi; Hassan, Yassin A

    2011-05-31

    An experimental activity was performed to observe and study the effects of graphite dispersion and deposition on thermal-hydraulic phenomena in a reactor cavity cooling system (RCCS). The small-scale RCCS experimental facility (16.5 x 16.5 x 30.4 cm) used for this activity represents half of the reactor cavity with an electrically heated vessel. Water flowing through five vertical pipes removes the heat produced in the vessel and releases it into the environment by mixing with cold water in a large tank. The particle image velocimetry technique was used to study the velocity field of the air inside the cavity. A set of 52 thermocouples was installed in the facility to monitor the temperature profiles of the vessel, pipe walls, and air. Ten grams of a fine graphite powder (average particle size 2 m) was injected into the cavity through a spraying nozzle placed at the bottom of the vessel. The temperatures and air velocity field were recorded and compared with the measurements obtained before the graphite dispersion, showing a decrease of the temperature surfaces that was related to an increase in their emissivity. The results contribute to the understanding of RCCS capability in an accident scenario.

  7. Commercial Light Water Reactor -Tritium Extraction Facility Process Waste Assessment (Project S-6091)

    Energy Technology Data Exchange (ETDEWEB)

    Hsu, R.H.; Delley, A.O.; Alexander, G.J.; Clark, E.A.; Holder, J.S.; Lutz, R.N.; Malstrom, R.A.; Nobles, B.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Carson, S.D. [Sandia National Laboratories, New Mexico, NM (United States); Peterson, P.K. [Sandia National Laboratories, New Mexico, NM (United States)

    1997-11-30

    The Savannah River Site (SRS) has been tasked by the Department of Energy (DOE) to design and construct a Tritium Extraction Facility (TEF) to process irradiated tritium producing burnable absorber rods (TPBARs) from a Commercial Light Water Reactor (CLWR). The plan is for the CLWR-TEF to provide tritium to the SRS Replacement Tritium Facility (RTF) in Building 233-H in support of DOE requirements. The CLWR-TEF is being designed to provide 3 kg of new tritium per year, from TPBARS and other sources of tritium (Ref. 1-4).The CLWR TPBAR concept is being developed by Pacific Northwest National Laboratory (PNNL). The TPBAR assemblies will be irradiated in a Commercial Utility light water nuclear reactor and transported to the SRS for tritium extraction and processing at the CLWR-TEF. A Conceptual Design Report for the CLWR-TEF Project was issued in July 1997 (Ref. 4).The scope of this Process Waste Assessment (PWA) will be limited to CLWR-TEF processing of CLWR irradiated TPBARs. Although the CLWR- TEF will also be designed to extract APT tritium-containing materials, they will be excluded at this time to facilitate timely development of this PWA. As with any process, CLWR-TEF waste stream characteristics will depend on process feedstock and contaminant sources. If irradiated APT tritium-containing materials are to be processed in the CLWR-TEF, this PWA should be revised to reflect the introduction of this contaminant source term.

  8. The DRAGON aerosol research facility to study aerosol behaviour for reactor safety applications

    Energy Technology Data Exchange (ETDEWEB)

    Suckow, Detlef; Guentay, Salih [Paul Scherrer Institut, Villigen (Switzerland)

    2008-07-01

    During a severe accident in a nuclear power plant fission products are expected to be released in form of aerosol particles and droplets. To study the behaviour of safety relevant reactor components under aerosol loads and prototypical severe accident conditions the multi-purpose aerosol generation facility DRAGON is used since 1994 for several projects. DRAGON can generate aerosol particles by the evaporation-condensation technique using a plasma torch system, fluidized bed and atomization of particles suspended in a liquid. Soluble, hygroscopic aerosol (i.e. CsOH) and insoluble aerosol particles (i.e. SnO{sub 2}, TiO{sub 2}) or mixtures of them can be used. DRAGON uses state-of-the-art thermal-hydraulic, data acquisition and aerosol measurement techniques and is mainly composed of a mixing chamber, the plasma torch system, a steam generator, nitrogen gas and compressed air delivery systems, several aerosol delivery piping, gas heaters and several auxiliary systems to provide vacuum, coolant and off-gas treatment. The facility can be operated at system pressure of 5 bars, temperatures of 300 deg. C, flow rates of non-condensable gas of 900 kg/h and steam of 270 kg/h, respectively. A test section under investigation is attached to DRAGON. The paper summarizes and demonstrates with the help of two project examples the capabilities of DRAGON for reactor safety studies. (authors)

  9. Summary of estimated doses and risks resulting from routine radionuclide releases from fast breeder reactor fuel cycle facilities

    Energy Technology Data Exchange (ETDEWEB)

    Miller, C.W.; Meyer, H.R.

    1985-01-01

    A project is underway at Oak Ridge National Laboratory to assess the human health and environment effects associated with operation of Liquid Metal Fast Breeder Reactor fuel cycle. In this first phase of the work, emphasis was focused on routine radionuclide releases from reactor and reprocessing facilities. For this study, sites for fifty 1-GW(e) capacity reactors and three reprocessing plants were selected to develop scenarios representative of US power requirements. For both the reactor and reprocessing facility siting schemes selected, relatively small impacts were calculated for locality-specific populations residing within 100 km. Also, the results of these analyses are being used in the identification of research priorities. 13 refs., 2 figs., 3 tabs.

  10. Experiments in the experimental fast reactor VENUS-F: The FREYA project; Experimentos en el reactor rapido experimental VENUS-F: El proyecto FREYA

    Energy Technology Data Exchange (ETDEWEB)

    Villamarin, D.; Becares, V.; Cano, D.; Gonzalez, E.

    2011-07-01

    Due to the high flexibility of operation of the reactor VENUS-E, FREYA project has two main objectives. The first is the end of the study monitoring techniques reactivity and serve as validation of simulation codes. The second objective is to provide experimental support for design and licensing MYRRHA / FASTEE and TRF in collaboration with CDTy LEADER projects of the 7th Framework Programme of the EU.

  11. Modernization of the facilities of the TRIGA Mark III reactor of ININ; Modernizacion de las instalaciones del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Mendez T, D.; Flores C, J., E-mail: dario.mendez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) has been in operation since 1968 under strict maintenance and component replacement programs, which has allowed its safe operation during this time. Under this scheme, the reactor was operating under suitable conditions, taking into account the different requests for operation that were received for the samples irradiation for the radioisotopes production such as the Sm-153, personnel training, basic research, archaeology and environmental studies and nuclear chemistry of the elements. However, a modernization program of its components and laboratories was required, in order to improve safety in the operation of the same and to increase its use in the analysis of samples by neutron activation and in the training of personnel. This program known as Modernization Program of the Reactor Facilities, was proposed alongside the project to replace high-enrichment fuels with low-enrichment fuels at the end of 2011 and early 2012. The central aspects of this program are described in this work, grouped into generic topics that include instrumentation and control, the radiological monitoring system of the area, the cooling system, the ventilation system, the neutron activation analysis laboratory, the manufacture of graphite elements, inspection submersible system of the pool, temporary storage system for irradiated fuels, traveling crane, Reactor support laboratories and technical meetings, courses and seminars for reactor personnel and associated groups. It also describes some of the most relevant components required for each system and the progress that is made in each one of them. As a fundamental result of the implementation of this Modernization Program of the Reactor Facilities, there has been a substantial improvement in the performance of the systems and components of its facilities, in the reliability of its operation and in the safety of the same. (Author)

  12. Experimental and computational investigation of flow of pebbles in a pebble bed nuclear reactor

    Science.gov (United States)

    Khane, Vaibhav B.

    The Pebble Bed Reactor (PBR) is a 4th generation nuclear reactor which is conceptually similar to moving bed reactors used in the chemical and petrochemical industries. In a PBR core, nuclear fuel in the form of pebbles moves slowly under the influence of gravity. Due to the dynamic nature of the core, a thorough understanding about slow and dense granular flow of pebbles is required from both a reactor safety and performance evaluation point of view. In this dissertation, a new integrated experimental and computational study of granular flow in a PBR has been performed. Continuous pebble re-circulation experimental set-up, mimicking flow of pebbles in a PBR, is designed and developed. Experimental investigation of the flow of pebbles in a mimicked test reactor was carried out for the first time using non-invasive radioactive particle tracking (RPT) and residence time distribution (RTD) techniques to measure the pebble trajectory, velocity, overall/zonal residence times, flow patterns etc. The tracer trajectory length and overall/zonal residence time is found to increase with change in pebble's initial seeding position from the center towards the wall of the test reactor. Overall and zonal average velocities of pebbles are found to decrease from the center towards the wall. Discrete element method (DEM) based simulations of test reactor geometry were also carried out using commercial code EDEM(TM) and simulation results were validated using the obtained benchmark experimental data. In addition, EDEM(TM) based parametric sensitivity study of interaction properties was carried out which suggests that static friction characteristics play an important role from a packed/pebble beds structural characterization point of view. To make the RPT technique viable for practical applications and to enhance its accuracy, a novel and dynamic technique for RPT calibration was designed and developed. Preliminary feasibility results suggest that it can be implemented as a non

  13. Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D., E-mail: thiagodbtr@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil); Silva, Mário A. B. da, E-mail: mabs500@gmail.com [Departamento de Energia Nuclear (CTG/UFPE), Av. Professor Luiz Freire, 1000, Recife 50740-540, PE (Brazil); Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil)

    2016-01-15

    The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

  14. Experimental Facilities Division progress report 1996--97

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    This progress report summarizes the activities of the Experimental Facilities Division (XFD) in support of the users of the Advanced Photon Source (APS), primarily focusing on the past year of operations. In September 1996, the APS began operations as a national user facility serving the US community of x-ray researchers from private industry, academic institutions, and other research organizations. The start of operations was about three months ahead of the baseline date established in 1988. This report is divided into the following sections: (1) overview; (2) user operations; (3) user administration and technical support; (4) R and D in support of view operations; (5) collaborative research; and (6) long-term strategic plans for XFD.

  15. Experimental investigation of a pilot-scale jet bubbling reactor for wet flue gas desulphurisation

    DEFF Research Database (Denmark)

    Zheng, Yuanjing; Kiil, Søren; Johnsson, Jan Erik

    2003-01-01

    In the present work, an experimental parameter study was conducted in a pilot-scale jet bubbling reactor for wet flue gas desulphurisation (FGD). The pilot plant is downscaled from a limestone-based, gypsum producing full-scale wet FGD plant. Important process parameters, such as slurry pH, inlet...... flue gas concentration of SO2, reactor temperature, and slurry concentration of Cl- have been varied. The degree of desulphurisation, residual limestone content of the gypsum, liquid phase concentrations, and solids content of the slurry were measured during the experimental series. The SO2 removal...

  16. Experimental studies on catalytic hydrogen recombiners for light water reactors; Experimentelle Untersuchungen zu katalytischen Wasserstoffkombinatoren fuer Leichtwasserreaktoren

    Energy Technology Data Exchange (ETDEWEB)

    Drinovac, P.

    2006-06-19

    In the course of core melt accidents in nuclear power plants a large amount of hydrogen can be produced and form an explosive or even detonative gas mixture with aerial oxygen in the reactor building. In the containment atmosphere of pressurized water reactors hydrogen combines a phlogistically with the oxygen present to form water vapor even at room temperature. In the past, experimental work conducted at various facilities has contributed little or nothing to an understanding of the operating principles of catalytic recombiners. Hence, the purpose of the present study was to conduct detailed investigations on a section of a recombiner essentially in order to deepen the understanding of reaction kinetics and heat transport processes. The results of the experiments presented in this dissertation form a large data base of measurements which provides an insight into the processes taking place in recombiners. The reaction-kinetic interpretation of the measured data confirms and deepens the diffusion theory - proposed in an earlier study. Thus it is now possible to validate detailed numeric models representing the processes in recombiners. Consequently the present study serves to broaden and corroborate competence in this significant area of reactor technology. In addition, the empirical knowledge thus gained may be used for a critical reassessment of previous numeric model calculations. (orig.)

  17. Analysis of radiological accident emissions of a lead-cooled experimental reactor. LEADER Project; Analisis radiologico de las emisiones en caso de accidente de un reactor experimental refrigerado por plomo. Proyecto LEADER

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Salcedo, F.; Cortes Martin, A.

    2013-07-01

    The LEADER project develops a conceptual level industrial size reactor cooled lead and a demonstration plant of this technology. The project objectives are to define the characteristics and design to installation scale reactor using available technologies and short-term components and assess safety aspects conducting a preliminary analysis of the impact of the facility.

  18. Design of Pump as Turbine Experimental Test Facility

    Directory of Open Access Journals (Sweden)

    Zariatin D. L.

    2017-01-01

    Full Text Available This paper presents the design process of experimental test facility for pump as turbine hydropower system. Three design possibilities that related to the PAT condition of operation was developed and analyzed by using CFD Software. It is found that the First Variant with a straight flow to the PAT will produce higher velocity, which is needed to generate more rotation of the shaft generator, in order to generate more electric power. The strength of PAT construction was analyzed by using FEM software. It was found that the maximum stress is 6 MPa and can be concluded that the construction is appropriate to the design requirement.

  19. Lewis Research Center's coal-fired, pressurized, fluidized-bed reactor test facility

    Science.gov (United States)

    Kobak, J. A.; Rollbuhler, R. J.

    1981-01-01

    A 200-kilowatt-thermal, pressurized, fluidized-bed (PFB) reactor, research test facility was designed, constructed, and operated as part of a NASA-funded project to assess and evaluate the effect of PFB hot-gas effluent on aircraft turbine engine materials that might have applications in stationary-power-plant turbogenerators. Some of the techniques and components developed for this PFB system are described. One of the more important items was the development of a two-in-one, gas-solids separator that removed 95+ percent of the solids in 1600 F to 1900 F gases. Another was a coal and sorbent feed and mixing system for injecting the fuel into the pressurized combustor. Also important were the controls and data-acquisition systems that enabled one person to operate the entire facility. The solid, liquid, and gas sub-systems all had problems that were solved over the 2-year operating time of the facility, which culminated in a 400-hour, hot-gas, turbine test.

  20. Experimental characterization of slurry bubble-column reactor hydrodynamics

    Energy Technology Data Exchange (ETDEWEB)

    Shollenberger, K.A.; Torczynski, J.R.; Jackson, N.B.; O`Hern, T.J.

    1997-09-01

    Sandia`s program to develop, implement, and apply diagnostics for hydrodynamic characterization of slurry bubble column reactors (SBCRs) at industrially relevant conditions is discussed. Gas liquid flow experiments are performed on an industrial scale. Gamma densitometry tomography (GDT) is applied to measure radial variations in gas holdup at one axial location. Differential pressure (DP) measurements are used to calculate volume averaged gas holdups along the axis of the vessel. The holdups obtained from DP show negligible axial variation for water but significant variations for oil, suggesting that the air water flow is fully developed (minimal flow variations in the axial direction) but that the air oil flow is still developing at the GDT measurement location. The GDT and DP gas holdup results are in good agreement for the air water flow but not for the air oil flow. Strong flow variations in the axial direction may be impacting the accuracy of one or both of these techniques. DP measurements are also acquired at high sampling frequencies (250 Hz) and are interpreted using statistical analyses to determine the physical mechanism producing each frequency component in the flow. This approach did not yield the information needed to determine the flow regime in these experiments. As a first step toward three phase material distribution measurements, electrical impedance tomography (EIT) and GDT are applied to a liquid solid flow to measure solids holdup. Good agreement is observed between both techniques and known values.

  1. Identification and Quantification of Carbon Phases in Conversion Fuel for the Transient Reactor Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Steele, Robert; Mata, Angelica; Dunzik-Gougar, Mary Lou; van Rooyen, Isabella

    2016-06-01

    As part of an overall effort to convert US research reactors to low-enriched uranium (LEU) fuel use, a LEU conversion fuel is being designed for the Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory. TREAT fuel compacts are comprised of UO2 fuel particles in a graphitic matrix material. In order to refine heat transfer modeling, as well as determine other physical and nuclear characteristics of the fuel, the amount and type of graphite and non-graphite phases within the fuel matrix must be known. In this study, we performed a series of complementary analyses, designed to allow detailed characterization of the graphite and phenolic resin based fuel matrix. Methods included Scanning Electron and Transmission Electron Microscopies, Raman spectroscopy, X-ray Diffraction, and Dual-Beam Focused Ion Beam Tomography. Our results indicate that no single characterization technique will yield all of the desired information; however, through the use of statistical and empirical data analysis, such as curve fitting, partial least squares regression, volume extrapolation and spectra peak ratios, a degree of certainty for the quantity of each phase can be obtained.

  2. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    Science.gov (United States)

    Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.

    2013-06-01

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  3. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  4. INDEPENDENT CONFIRMATORY SURVEY REPORT FOR THE REACTOR BUILDING, HOT LABORATORY, PRIMARY PUMP HOUSE, AND LAND AREAS AT THE PLUM BROOK REACTOR FACILITY, SANDUSKY, OHIO

    Energy Technology Data Exchange (ETDEWEB)

    Erika N. Bailey

    2011-10-10

    In 1941, the War Department acquired approximately 9,000 acres of land near Sandusky, Ohio and constructed a munitions plant. The Plum Brook Ordnance Works Plant produced munitions, such as TNT, until the end of World War II. Following the war, the land remained idle until the National Advisory Committee for Aeronautics later called the National Aeronautics and Space Administration (NASA) obtained 500 acres to construct a nuclear research reactor designed to study the effects of radiation on materials used in space flight. The research reactor was put into operation in 1961 and was the first of fifteen test facilities eventually built by NASA at the Plum Brook Station. By 1963, NASA had acquired the remaining land at Plum Brook for these additional test facilities

  5. Comparisons of RELAP5-3D Analyses to Experimental Data from the Natural Convection Shutdown Heat Removal Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Hu, Rui; Lisowski, Darius; Kraus, Adam

    2016-04-17

    The Reactor Cavity Cooling System (RCCS) is an important passive safety system being incorporated into the overall safety strategy for high temperature advanced reactor concepts such as the High Temperature Gas- Cooled Reactors (HTGR). The Natural Convection Shutdown Heat Removal Test Facility (NSTF) at Argonne National Laboratory (Argonne) reflects a 1/2-scale model of the primary features of one conceptual air-cooled RCCS design. The project conducts ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy’s Advanced Reactor Technology (ART) program, while also generating data for code validation purposes. While experiments are being conducted at the NSTF to evaluate the feasibility of the passive RCCS, parallel modeling and simulation efforts are ongoing to support the design, fabrication, and operation of these natural convection systems. Both system-level and high fidelity computational fluid dynamics (CFD) analyses were performed to gain a complete understanding of the complex flow and heat transfer phenomena in natural convection systems. This paper provides a summary of the RELAP5-3D NSTF model development efforts and provides comparisons between simulation results and experimental data from the NSTF. Overall, the simulation results compared favorably to the experimental data, however, further analyses need to be conducted to investigate any identified differences.

  6. Design of a management information system for the Shielding Experimental Reactor ageing management

    Energy Technology Data Exchange (ETDEWEB)

    He Jie, E-mail: hejiejoe@163.co [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Xu Xianhong [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2010-01-15

    The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

  7. Antenna design for fast ion collective Thomson scattering diagnostic for the international thermonuclear experimental reactor

    DEFF Research Database (Denmark)

    Leipold, Frank; Furtula, Vedran; Salewski, Mirko

    2009-01-01

    Fast ion physics will play an important role for the international thermonuclear experimental reactor (ITER), where confined alpha particles will affect and be affected by plasma dynamics and thereby have impacts on the overall confinement. A fast ion collective Thomson scattering (CTS) diagnostic...

  8. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  9. A scaled experimental study of control blade insertion dynamics in Pebble-Bed Fluoride-Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Buster, Grant C., E-mail: grant.buster@gmail.com; Laufer, Michael R.; Peterson, Per F.

    2016-07-15

    Highlights: • A granular dynamics scaling methodology is discussed. • Control blade insertion in a representative pebble-bed core is experimentally studied. • Control blade insertion forces and pebble displacements are experimentally measured. • X-ray tomography techniques are used to observe pebble displacement distributions. - Abstract: Direct control element insertion into a pebble-bed reactor core is proposed as a viable control system in molten-salt-cooled pebble-bed reactors. Unlike helium-cooled pebble-bed reactors, this reactor type uses spherical fuel elements with near-neutral buoyancy in the molten-salt coolant, thus reducing contact forces on the fuel elements. This study uses the X-ray Pebble Bed Recirculation Experiment facility to measure the force required to insert a control element directly into a scaled pebble-bed. The required control element insertion force, and therefore the contact force on fuel elements, is measured to be well below recommended limits. Additionally, X-ray tomography is used to observe how the direct insertion of a control element physically displaces spherical fuel elements. The tomography results further support the viability of direct control element insertion into molten-salt-cooled pebble-bed reactor cores.

  10. IAEA coordinated research program on `harmonization and validation of fast reactor thermomechanical and thermohydraulic codes using experimental data`. 1. Thermohydraulic benchmark analysis on high-cycle thermal fatigue events occurred at French fast breeder reactor Phenix

    Energy Technology Data Exchange (ETDEWEB)

    Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1997-06-01

    A benchmark exercise on `Tee junction of Liquid Metal Fast Reactor (LMFR) secondary circuit` was proposed by France in the scope of the said Coordinated Research Program (CRP) via International Atomic Energy Agency (IAEA). The physical phenomenon chosen here deals with the mixture of two flows of different temperature. In a LMFR, several areas of the reactor are submitted to this problem. They are often difficult to design, because of the complexity of the phenomena involved. This is one of the major problems of the LMFRs. This problem has been encountered in the Phenix reactor on the secondary loop, where defects in a tee junction zone were detected during a campaign of inspections after an operation of 90,000 hours of the reactor. The present benchmark is based on an industrial problem and deal with thermal striping phenomena. Problems on pipes induced by thermal striping phenomena have been observed in some reactors and experimental facilities coolant circuits. This report presents numerical results on thermohydraulic characteristics of the benchmark problem, carried out using a direct numerical simulation code DINUS-3 and a boundary element code BEMSET. From the analysis with both the codes, it was confirmed that the hot sodium from the small pipe rise into the cold sodium of the main pipe with thermally instabilities. Furthermore, it was indicated that the coolant mixing region including the instabilities agrees approximately with the result by eye inspections. (author)

  11. The manufacture of enriched uranium fuel slugs for the Experimental Breeder Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Shuck, Author B.

    1953-04-20

    This report describes the specifications, materials and the sequence of operations used to found and fabricate 4 the first charge of enriched uranium fuel in the Experimental Breeder Reactor. The work was governed by the following principles: a. That the fuel be of correct composition, dimension and metallurgical condition for use in the reactor. b. That a maximum yield of finished fuel slugs from the quantity of uranium available for the program be achieved. c. That the residues be in a form which can be recovered by chemical or other means. d. That a detailed record be kept in such form that a complete history of each fuel slug be available.

  12. Management of waste from the International Thermonuclear Experimental Reactor and from future fusion power plants

    Energy Technology Data Exchange (ETDEWEB)

    Broden, K. [Association EURATOM, Nykoeping (Sweden); Lindberg, M. [Association EURATOM, Nykoeping (Sweden); Nisan, S. [The NET Team, Garching (Germany); Rocco, P. [European Commission, Institute for Advanced Materials, Joint Research Centre, Ispra (Vatican City State, Holy See) (Italy); Zucchetti, M. [Energetics Department, Polytechnic of Turin, Corso Duca degli Abruzzi 24, I-10129 Torino (Italy); Taylor, N. [Association EURATOM-UKAEA, UKAEA Fusion, Culham, Abingdon, Oxfordshire, OX14 3DB (United Kingdom); Forty, C. [Association EURATOM-UKAEA, UKAEA Fusion, Culham, Abingdon, Oxfordshire, OX14 3DB (United Kingdom)

    1997-04-01

    An important inherent advantage of fusion would be the total absence of high-level radioactive spent fuel as produced in fission reactors. Fusion will, however, produce activated material containing both activation products and tritium. Part of the material may also contain chemically toxic substances. This paper describes methods that could be used to manage these materials and also methods to reduce or entirely eliminate the waste quantities. The results are based on studies for the International Thermonuclear Experimental Reactor and also for future fusion power station designs currently under investigation within the European programme on the safety and environmental assessment of fusion power, long-term. (orig.)

  13. SBLOCA AND LOFW EXPERIMENTS IN A SCALED-DOWN IET FACILITY OF REX-10 REACTOR

    Directory of Open Access Journals (Sweden)

    YEON-GUN LEE

    2013-06-01

    Full Text Available This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA and the loss-of-feedwater accident (LOFW in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF, a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

  14. Experimental studies into the thermal-hydraulic performance of the VK-300 reactor based on a draft tube model

    Directory of Open Access Journals (Sweden)

    N.P. Serdun

    2015-12-01

    Full Text Available The paper presents an experimental study into the thermal-hydraulic performance of the VK-300 reactor based on a model of a single draft tube at a pressure of 3.4MPa, various flow rates and the model inlet relative enthalpies of –0.05 to 0.2. The experimental procedures include generation of a steam-water mixture circulation with a preset flow rate and a relative enthalpy through the test section at a pressure of 3.3 to 3.4MPa, and measurement of thermal-hydraulic parameters within the circuit's representative upflow and downflow lengths of practical interest. There have been confirmed the designs used to support the reactor facility serviceability and the assumptions concerning the thermal-hydraulic performance of a natural circulation circuit used in the analysis thereof. It has been shown that, across the analyzed range of the relative enthalpy values, the draft tube has an annular-dispersed or an annular flow of the steam-water mixture, both providing for the significant separation of the steam-water mixture (Ksep=0.4 at the draft tube edges and in the mixing chamber. The perforation in the upper part of the draft tubes allows the separation coefficient to be increased at the first stage and creates more favorable conditions for the second-stage separation. The measured values of the void fraction in the mixing chamber and in the draft tube are in a satisfactory agreement with calculations based on Z.L. Miropolskiy's method and the RELAP code and may be used to verify the VK-300 thermal-hydraulic codes. It has been shown that steam may enter the ring slit that simulates the annular space and reach the reactor core inlet. Further investigations need to be conducted to study this effect for its guaranteed exclusion and for the development of emergency response procedures.

  15. Geothermal Loop Experimental Facility. Quarterly report, April--June 1978

    Energy Technology Data Exchange (ETDEWEB)

    Bischoff, W.S.; Bishop, H.K.; Cooney, C.S.; Hanenburg, W.H.; Hoaglin, G.J.; Jacobson, W.O.; Mulliner, D.K.; Newell, D.G.; Swanson, C.R.

    1978-07-01

    The Geothermal Loop Experimental Facility (GLEF) was modified to use a two stage flash process with two parallel flash trains for the extraction of energy from a high temperature, high salinity, liquid-dominated resource. Since plant start-up in May 1976, a substantial amount of information has been obtained on the operation of the plant, components, brine and steam composition, production and injection wells, and the potential of the Niland Reservoir. The general operation and accomplishments of the GLEF during the period April 1978 through June 1978 are discussed. The GLEF underwent a major redesign. Modifications and inspections of various GLEF equipment and systems are also discussed. Information about the production and injection wells flow testing and instrumentation are discussed. Information regarding coatings and linings for valves and piping is included. In the Chemistry Section there is a wide range of data taken from Brine, Steam, Scale, Binary, Condensate, and Cooling Water Systems.

  16. Experimental facility for testing nuclear instruments for planetary landing missions

    Science.gov (United States)

    Golovin, Dmitry; Mitrofanov, Igor; Litvak, Maxim; Kozyrev, Alexander; Sanin, Anton; Vostrukhin, Andrey

    2017-04-01

    The experimental facility for testing and calibration of nuclear planetology instruments has been built in the frame of JINR and Space Research Institute (Moscow) cooperation. The Martian soil model from silicate glass with dimensions 3.82 x 3.21 m and total weight near 30 tons has been assembled in the facility. The glass material was chosen for imitation of dry Martian regolith. The heterogeneous model has been proposed and developed to achieve the most possible similarity with Martian soil in part of the average elemental composition by adding layers of necessary materials, such as iron, aluminum, and chlorine. The presence of subsurface water ice is simulated by adding layers of polyethylene at different depths inside glass model assembly. Neutron generator was used as a neutron source to induce characteristic gamma rays for testing active neutron and gamma spectrometers to define elements composition of the model. The instrumentation was able to detect gamma lines attributed to H, O, Na, Mg, Al, Si, Cl, K, Ca and Fe. The identified elements compose up to 95 wt % of total mass of the planetary soil model. This results will be used for designing scientific instruments to performing experiments of active neutron and gamma ray spectroscopy on the surface of the planets during Russian and international missions Luna-Glob, Luna-Resource and ExoMars-2020.

  17. On the possibility of experimentally confirming the hypothesis of reactor antineutrino passage into a sterile state

    Science.gov (United States)

    Serebrov, A. P.; Fomin, A. K.; Zinov'ev, V. G.; Loginov, Yu. E.; Onegin, M. S.; Gagarskiy, A. M.; Petrov, G. A.; Solovei, V. A.; Chernyi, A. V.; Zherebtsov, O. M.; Martem'yanov, V. P.; Tsinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Svyatkin, M. N.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Khramkov, N. S.; Rykalin, V. I.

    2013-07-01

    The "Neutrino-4" experiment for the 100-MW SM-3 reactor has been developed with the aim of testing the reactor antineutrino anomaly at Petersburg Nuclear Physics Institute. The advantages of this reactor for studying the antineutrino anomaly are (i) a low background level and (ii) small dimensions (35 × 42 × 42 cm) of the active zone. Operation of a position-sensitive antineutrino detector comprising five working sections and moving so as to cover a region of distances within 6-13 m from the active zone has been simulated by the Monte-Carlo method. The range of experimental sensitivity with respect to the oscillation parameters Δ m 2 and sin22θ is determined, which will make it possible to confirm the hypothesis of antineutrino oscillations into a sterile state.

  18. Material unaccounted for at the Southwest Experimental Fast Oxide Reactor: The SEFOR MUF

    Energy Technology Data Exchange (ETDEWEB)

    Higinbotham, W.A.

    1994-11-07

    The U.S. Atomic Energy Commission contracted with the General Electric Company to design, construct, and operate the Southwest Experimental Fast Oxide Reactor (SEFOR) to measure the Doppler effect for fast neutron breeder reactors. It contracted with Nuclear Fuel Services to fabricate the fuel rods for the reactor. When the reactor went critical in May, 1969, it appeared that some of the mixed uranium-plutonium oxide (MOX) fuel rods did not contain the specified quantity of plutonium. The SEFOR operators soon found several fuel rods which appeared to be low in plutonium. The safeguards group at Brookhaven was asked to look into the problem and, if possible, determine how much plutonium was missing from the unirradiated rods and from the larger number which had been slightly irradiated in the reactor. It was decided that the plutonium content of the unirradiated and irradiated rods could be measured relative to a reference rod using a high resolution gamma-ray detector and also by neutron measurements using an auto-correlation circuit recently developed at the Naval Research Laboratory (NRL). During the next two years, Brookhaven personnel and C.V. Strain of NRL made several trips to the SEFOR reactor. About 250 of the 775 rods were measured by two or more methods, using a sodium-iodide detector, a high-resolution germanium detector, a neutron detector, or the reactor (to measure reactivity). The research team concluded that 4.6 {+-} 0.46 kg of plutonium was missing out of the 433 kg that the rods should have contained. This report describes the SEFOR experiment and the procedures used to determine the material unaccounted for, or MUF.

  19. COMPUTATIONAL AND EXPERIMENTAL MODELING OF SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Paul Lam; Dimitri Gidaspow

    2000-09-01

    The objective if this study was to develop a predictive experimentally verified computational fluid dynamics (CFD) model for gas-liquid-solid flow. A three dimensional transient computer code for the coupled Navier-Stokes equations for each phase was developed. The principal input into the model is the viscosity of the particulate phase which was determined from a measurement of the random kinetic energy of the 800 micron glass beads and a Brookfield viscometer. The computed time averaged particle velocities and concentrations agree with PIV measurements of velocities and concentrations, obtained using a combination of gamma-ray and X-ray densitometers, in a slurry bubble column, operated in the bubbly-coalesced fluidization regime with continuous flow of water. Both the experiment and the simulation show a down-flow of particles in the center of the column and up-flow near the walls and nearly uniform particle concentration. Normal and shear Reynolds stresses were constructed from the computed instantaneous particle velocities. The PIV measurement and the simulation produced instantaneous particle velocities. The PIV measurement and the simulation produced similar nearly flat horizontal profiles of turbulent kinetic energy of particles. This phase of the work was presented at the Chemical Reaction Engineering VIII: Computational Fluid Dynamics, August 6-11, 2000 in Quebec City, Canada. To understand turbulence in risers, measurements were done in the IIT riser with 530 micron glass beads using a PIV technique. The results together with simulations will be presented at the annual meeting of AIChE in November 2000.

  20. Modeling of Heat Transfer in the Helical-Coil Heat Exchanger for the Reactor Facility "UNITERM"

    Directory of Open Access Journals (Sweden)

    V. I. Solonin

    2014-01-01

    Full Text Available Circuit heat sink plays an important role in the reactor system. Therefore it imposes high requirements for quality of determining thermal-hydraulic parameters. This article is aimed at modeling of heat exchange process of the helical-coil heat exchanger, which is part of the heat sink circuit of the reactor facility "UNITERM."The simulation was performed using hydro-gas-dynamic software package ANSYS CFX. Computational fluid dynamics of this package allows us to perform calculations in a threedimensional setting, giving an idea of the fluid flow nature. The purpose of the simulation was to determine the parameters of the helical-coil heat exchanger (temperature, velocity at the outlet of the pipe and inter-tubular space, pressure drop, and the nature of the fluid flow of primary and intermediate coolants. Geometric parameters of the model were determined using the preliminary calculations performed by the criterion equations. In calculations Turbulence models k-ε RNG, Shear Stress Transport (SST are used. The article describes selected turbulence models, and considers relationship with wall function.The calculation results allow us to give the values obtained for thermal-hydraulic parameters, to compare selected turbulence models, as well as to show distribution patterns of the coolant temperature, pressure, and velocity at the outlet of the intermediate cooler.Calculations have shown that:- maximum values of primary coolant temperature at the outlet of the heat exchanger surface are encountered in the space between the helical-coil tubes;- higher temperatures of intermediate coolant at the outlet of the coils (in space of helicalcoil tubes are observed for the peripheral row;- primary coolant movement in the inter-tubular space of helical-coil surface is formed as a spiral flow, rather than as a in-line tube bank cross flow.

  1. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

  2. Experimental and Kinetic Modeling Study of Ethyl Levulinate Oxidation in a Jet-Stirred Reactor

    KAUST Repository

    Wang, Jui-Yang

    2017-06-01

    A jet-stirred reactor was designed and constructed in the Clean Combustion Research Center (CCRC) at King Abdullah University of Science and Technology (KAUST); was validated with n-heptane, iso-octane oxidation and cyclohexene pyrolysis. Different configurations of the setup have been tested to achieve good agreement with results from the literature. Test results of the reactor indicated that installation of a pumping system at the downstream side in the experimental apparatus was necessary to avoid the reoccurrence of reactions in the sampling probe. Experiments in ethyl levulinate oxidation were conducted in the reactor under several equivalence ratios, from 600 to 1000 K, 1 bar and 2 s residence time. Oxygenated species detected included methyl vinyl ketone, levulinic acid and ethyl acrylate. Ethylene, methane, carbon monoxide, hydrogen, oxygen and carbon dioxide were further quantified with a gas chromatography, coupled with a flame ionization detector and a thermal conductivity detector. The ethyl levulinate chemical kinetic model was first developed by Dr. Stephen Dooley, Trinity College Dublin, and simulated under the same conditions, using the Perfect-Stirred Reactor code in Chemkin software. In comparing the simulation results with experimental data, some discrepancies were noted; predictions of ethylene production were not well matched. The kinetic model was improved by updating several classes of reactions: unimolecular decomposition, H-abstraction, C-C and C-O beta-scissions of fuel radicals. The updated model was then compared again with experimental results and good agreement was achieved, proving that the concerted eliminated reaction is crucial for the kinetic mechanism formulation of ethyl levulinate. In addition, primary reaction pathways and sensitivity analysis were performed to describe the role of molecular structure in combustion (800 and 1000 K for ethyl levulinate oxidation in the jet-stirred reactor).

  3. Results of theoretical and experimental studies of hydrodynamics of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors

    Science.gov (United States)

    Ryabov, G. A.; Folomeev, O. M.; Sankin, D. A.; Melnikov, D. A.

    2015-02-01

    Problems of the calculation of circulation loops in circulating fluidized bed reactors and systems with interconnected reactors (polygeneration systems for the production of electricity, heat, and useful products and chemical cycles of combustion and gasification of solid fuels)are considered. A method has been developed for the calculation of circulation loop of fuel particles with respect to boilers with circulating fluidized bed (CFB) and systems with interconnected reactors with fluidized bed (FB) and CFB. New dependences for the connection between the fluidizing agent flow (air, gas, and steam) and performance of reactors and for the whole system (solids flow rate, furnace and cyclone pressure drops, and bed level in the riser) are important elements of this method. Experimental studies of hydrodynamics of circulation loops on the aerodynamic unit have been conducted. Experimental values of pressure drop of the horizontal part of the L-valve, which satisfy the calculated dependence, have been obtained.

  4. Experimentation on the anaerobic filter reactor for biogas production using rural domestic wastewater

    Science.gov (United States)

    Leju Celestino Ladu, John; Lü, Xi-wu; Zhong, Zhaoping

    2017-08-01

    The biogas production from anaerobic filter (AF) reactor was experimented in Taihu Lake Environmental Engineering Research Center of Southeast University, Wuxi, China. Two rounds of experimental operations were conducted in a laboratory scale at different Hydraulic retention time (HRT) and wastewater temperature. The biogas production rate during the experimentation was in the range of 4.63 to 11.78 L/d. In the first experimentation, the average gas production rate was 10.08 L/d, and in the second experimentation, the average gas production rate was 4.97 L/d. The experimentation observed the favorable Hydraulic Retention Time and wastewater temperature in AF was three days and 30.95°C which produced the gas concentration of 11.78 L/d. The HRT and wastewater temperature affected the efficiency of the AF process on the organic matter removal and nutrients removal as well. It can be deduced from the obtained results that HRT and wastewater temperature directly affects the efficiency of the AF reactor in biogas production. In conclusion, anaerobic filter treatment of organic matter substrates from the rural domestic wastewater increases the efficiency of the AF reactor on biogas production and gives a number of benefits for the management of organic wastes as well as reduction in water pollution. Hence, the operation of the AF reactor in rural domestic wastewater treatment can play an important element for corporate economy of the biogas plant, socio-economic aspects and in the development of effective and feasible concepts for wastewater management, especially for people in rural low-income areas.

  5. Completion Summary for Well NRF-16 near the Naval Reactors Facility, Idaho National Laboratory, Idaho

    Science.gov (United States)

    Twining, Brian V.; Fisher, Jason C.; Bartholomay, Roy C.

    2010-01-01

    In 2009, the U.S. Geological Survey in cooperation with the U.S. Department of Energy's Naval Reactors Laboratory Field Office, Idaho Branch Office cored and completed well NRF-16 for monitoring the eastern Snake River Plain (SRP) aquifer. The borehole was initially cored to a depth of 425 feet below land surface and water samples and geophysical data were collected and analyzed to determine if well NRF-16 would meet criteria requested by Naval Reactors Facility (NRF) for a new upgradient well. Final construction continued after initial water samples and geophysical data indicated that NRF-16 would produce chemical concentrations representative of upgradient aquifer water not influenced by NRF facility disposal, and that the well was capable of producing sustainable discharge for ongoing monitoring. The borehole was reamed and constructed as a Comprehensive Environmental Response Compensation and Liability Act monitoring well complete with screen and dedicated pump. Geophysical and borehole video logs were collected after coring and final completion of the monitoring well. Geophysical logs were examined in conjunction with the borehole core to identify primary flow paths for groundwater, which are believed to occur in the intervals of fractured and vesicular basalt and to describe borehole lithology in detail. Geophysical data also were examined to look for evidence of perched water and the extent of the annular seal after cement grouting the casing in place. Borehole videos were collected to confirm that no perched water was present and to examine the borehole before and after setting the screen in well NRF-16. Two consecutive single-well aquifer tests to define hydraulic characteristics for well NRF-16 were conducted in the eastern SRP aquifer. Transmissivity and hydraulic conductivity averaged from the aquifer tests were 4.8 x 103 ft2/d and 9.9 ft/d, respectively. The transmissivity for well NRF-16 was within the range of values determined from past aquifer

  6. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  7. Kinetics of vinyl acetate emulsion polymerization in a pulsed tubular reactor: comparison between experimental and simulation results

    OpenAIRE

    Sayer C.; Palma M.; Giudici R.

    2002-01-01

    A new reactor, the pulsed sieve plate column (PSPC), was developed to perform continuous emulsion polymerization reactions. This reactor combines the enhanced flexibility of tubular reactors with the mixing behavior provided by sieved plates and by the introduction of pulses that is important to prevent emulsion destabilization. The main objective of this work is to study the kinetics of vinyl acetate (VA) emulsion polymerization reactions performed in this PSPC. Therefore, both experimental ...

  8. High energy resolution characteristics on 14MeV neutron spectrometer for fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Iguchi, Tetsuo [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab.; Takada, Eiji; Nakazawa, Masaharu

    1996-10-01

    A 14MeV neutron spectrometer suitable for an ITER-like fusion experimental reactor is now under development on the basis of a recoil proton counter telescope principle in oblique scattering geometry. To verify its high energy resolution characteristics, preliminary experiments are made for a prototypical detector system. The comparison results show reasonably good agreement and demonstrate the possibility of energy resolution of 2.5% in full width at half maximum for 14MeV neutron spectrometry. (author)

  9. Simulation of Collimator for Neutron Imaging Facility of TRIGA MARK II PUSPATI Reactor

    Science.gov (United States)

    Zin, Muhammad Rawi Mohamed; Jamro, Rafhayudi; Yazid, Khairiah; Hussain, Hishamuddin; Yazid, Hafizal; Ahmad, Megat Harun Al Rashid Megat; Azman, Azraf; Mohamad, Glam Hadzir Patai; Hamzah, Nai'im Syaugi; Abu, Mohamad Puad

    Neutron Radiography facility in TRIGA MARK II PUSPATI reactor is being upgraded to obtain better image resolution as well as reducing exposure time. Collimator and exposure room are the main components have been designed for fabrication. This article focuses on the simulation part that was carried out to obtain the profile of collimated neutron beam by utilizing the neutron transport protocol code in the Monte Carlo N-Particle (MCNP) software. Particular interest is in the selection of materials for inlet section of the collimator. Results from the simulation indicates that a combination of Bismuth and Sapphire, each of which has 5.0 cm length that can significantly filter both the gamma radiation and the fast neutrons. An aperture made of Cadmium with 1.0 cm opening diameter provides thermal neutron flux about 1.8 x108 ncm-2s-1 at the inlet, but reduces to 2.7 x106 ncm-2s-1 at the sample plane. Still the flux obtained is expected to reduces exposure time as well as gaining better image resolution.

  10. The fast-spectrum transmutation experimental facility FASTEF: Main design achievements (Part 1: Core and primary system) within the FP7-CDT collaborative project of the European Commission

    Energy Technology Data Exchange (ETDEWEB)

    De Bruyn, D.; Fernandez, R. [Belgian Nuclear Research Centre (SCK CEN), Boeretang 200, 2400 Mol (Belgium); Mansani, L. [ANSALDO, Corso Perrone 25, 16152 Genova (Italy); Woaye-Hune, A. [AREVA-NP, rue Juliette Recamier 10, 69456 Lyon Cedex 06 (France); Sarotto, M. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Bubelis, E. [KIT, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2012-07-01

    MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK CEN in replacement of its material testing reactor BR2. SCK CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of three years. In this paper, we present the latest configuration of the reactor core and primary system. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1 2. If it remains a small-scale facility, the core power amounts now up to 100 MWth in critical mode. In a companion paper 3, we present the concept of the reactor building and the plant layout. (authors)

  11. An experimental study of hypervapotron structure in external reactor vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

    2016-07-15

    Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

  12. Development of a membrane-assisted fluidized bed reactor - 2 - Experimental demonstration and modeling for the partial oxidation of methanol

    NARCIS (Netherlands)

    Deshmukh, S.A.R.K.; Laverman, J.A.; van Sint Annaland, M.; Kuipers, J.A.M.

    2005-01-01

    A small laboratory-scale membrane-assisted fluidized bed reactor (MAFBR) was constructed in order to experimentally demonstrate the reactor concept for the partial oxidation of methanol to formaldehyde. Methanol conversion and product selectivities were measured at various overall fluidization

  13. Experimental and Numerical Evaluation of the By-Pass Flow in a Catalytic Plate Reactor for Hydrogen Production

    DEFF Research Database (Denmark)

    Sigurdsson, Haftor Örn; Kær, Søren Knudsen

    2011-01-01

    Numerical and experimental study is performed to evaluate the reactant by-pass flow in a catalytic plate reactor with a coated wire mesh catalyst for steam reforming of methane for hydrogen generation. By-pass of unconverted methane is evaluated under different wire mesh catalyst width to reactor...

  14. Development of Aerosol Measurement, Sampling and Generation Experimental Facilities under High Temperature and High Pressure

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jong Chan; Jung, Woo Young; Lee, Hyun Chul; Lee, Doo Yong [FNC TECH., Yongin (Korea, Republic of)

    2016-05-15

    The pressure is key factor determining Knudsen number and it affects aerosol dynamics. Hence, it is important to design experimental facilities to simulate the aerosols transport phenomena and removal mechanisms in the high temperature and high pressure (HT/HP). In case of Nuclear Power Plant, during the Light Water Reactor (LWR) severe accident, core degradation results in the release of both vapors and aerosol particles which differ in composition depending on their source terms. Vapor and aerosols generated under severe accident enter the containment atmosphere and are distributed in the containment by atmospheric flow. Temperature and pressure in the containment increase until containment spray system, fan cooler system or FCVS initiates to extract heat and avoid pressurization. The main purpose of the experimental facility is to develop not only multi-purpose test loops applying for aerosol industry but also to evaluation performance of engineered safety system including containment filtered venting system. The main experiment will be carried out in this loop and provide representative behavior of the aerosols under HT/HP conditions. The aim of the research is to be able to 1) develop the aerosol generation, mixing, sampling and measurement system and conduct tests based on various aerosol concentration, thermal-hydraulic conditions including high temperature and pressure and type of carrier gases (air, nitrogen and steam), applicable to the thermal power plant, environmental industry, automobile exhaust gas, chemical plant, HVAC system including nuclear power plant, and 2) investigate aerosol behaviors and removal mechanisms under these conditions. The tests with the main carrier gas of air will be performed on PHASEⅠ, steam will be conducted on PHASEⅡ.

  15. Reactor Safety Research Programs

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-07-01

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

  16. Plutonium Consumption Program, CANDU Reactor Project: Feasibility of BNFP Site as MOX Fuel Supply Facility. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-06-30

    An evaluation was made of the technical feasibility, cost, and schedule for converting the existing unused Barnwell Nuclear Fuel Facility (BNFP) into a Mixed Oxide (MOX) CANDU fuel fabrication plant for disposition of excess weapons plutonium. This MOX fuel would be transported to Ontario where it would generate electricity in the Bruce CANDU reactors. Because CANDU MOX fuel operates at lower thermal load than natural uranium fuel, the MOX program can be licensed by AECB within 4.5 years, and actual Pu disposition in the Bruce reactors can begin in 2001. Ontario Hydro will have to be involved in the entire program. Cost is compared between BNFP and FMEF at Hanford for converting to a CANDU MOX facility.

  17. Conceptual studies of toroidal field magnets for the tokamak experimental power reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Buncher, B.R.; Chi, J.W.H.; Fernandez, R.

    1976-10-26

    This report documents the principal results of a Conceptual Design Study for the Superconducting Toroidal Field System for a Tokamak Experimental Power Reactor. Two concepts are described for peak operating fields at the windings of 8 tesla, and 12 tesla, respectively. The design and manufacturing considerations are treated in sufficient detail that cost and schedule estimates could be developed. Major uncertainties in the design are identified and their potential impact discussed, along with recommendations for the necessary research and development programs to minimize these uncertainties. The minimum dimensions of a sub-size test coil for experimental qualification of the full size design are developed and a test program is recommended.

  18. Experimental investigation of in-vessel mixing phenomena in a VVER-1000 scaled test facility during unsteady asymmetric transients

    Energy Technology Data Exchange (ETDEWEB)

    Bucalossi, A. [EC JRC, (JRC F.5) PO Box 2, 1755 ZG Petten (Netherlands); Moretti, F.; Melideo, D. [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (UNIPI) 2, via Diotisalvi, 56100 Pisa (Italy); Del Nevo, A., E-mail: delnevo@hotmail.com [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (UNIPI) 2, via Diotisalvi, 56100 Pisa (Italy); D' Auria, F. [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Universita di Pisa (UNIPI) 2, via Diotisalvi, 56100 Pisa (Italy); Hoehne, T. [Forschungszentrum Dresden-Rossendorf (FZD), P.O.B. 51 01 19, D-01314 Dresden (Germany); Lisenkov, E. [FSUE OKB Gidropress, Ordshonikidize 21, RU-142103 Podolsk, Moscow district (Russian Federation); Gallori, D. [AREVA NP SAS, Tour AREVA - 92084 Paris, La Defense Cedex (France)

    2011-08-15

    Highlights: > Five mixing experiments in a scaled model of a VVER-1000 are described and discussed. > In-vessel mixing investigations of the coolant properties distribution at the core inlet. > These tests brought an improvement to existing experimental database for TH code validation. - Abstract: In-vessel turbulent mixing phenomena affect the time and space distribution of coolant properties (e.g., boron concentration and temperature) at the core inlet which impacts consequently the neutron kinetics response. For reactor safety evaluation purposes and to characterize these phenomena it is necessary to set and validate appropriate numerical modelling tools to improve the current conservative predictions. With such purpose, an experimental campaign was carried out by OKB Gidropress, in the framework of the European Commission Project 'TACIS R2.02/02 - Development of safety analysis capabilities for VVER-1000 transients involving spatial variations of coolant properties (temperature or boron concentration) at core inlet'. The experiments were conducted on a scaled facility representing the primary system of a VVER-1000 including a detailed model of the Reactor Pressure Vessel with its internals. The simulated transients involved perturbations of coolant properties distribution providing a wide validation matrix. The main achievements of the set of experiments featuring transient asymmetric pump behaviour are presented in this paper. The potential of the obtained experimental database for the validation of thermal fluid dynamics numerical simulation tools is also discussed and the role of computational fluid dynamics in supporting the experimental data analysis is highlighted.

  19. Present Status and Future Plans of J-PARC Hadron Experimental Facility

    CERN Document Server

    Tanaka, K

    2015-01-01

    Recovery of J-PARC Hadron Experimental Facility from the radioactive material leakage incident occurred on May 23, 2013 is reported. Recovery took long time. However its essential part was completed by the beginning of Japanese Fiscal Year 2015. Then we could start the beam operation of Hadron Experimental Facility from April 9, 2015. Experiments with slow extraction beam started on April 24, 2015. The beam intensity delivered to Hadron Experimental Facility reached approximately 32kW by the end of June, 2015. Recent activities on partic le and nuclear physics in the Hadron Experimental Facility are described also.

  20. Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014

    Energy Technology Data Exchange (ETDEWEB)

    Soelberg, Renae [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-11-01

    Advanced Test Reactor National Scientific User Facility (ATR NSUF) Monthly Report November 2014 Highlights Rory Kennedy and Sarah Robertson attended the American Nuclear Society Winter Meeting and Nuclear Technology Expo in Anaheim, California, Nov. 10-13. ATR NSUF exhibited at the technology expo where hundreds of meeting participants had an opportunity to learn more about ATR NSUF. Dr. Kennedy briefed the Nuclear Engineering Department Heads Organization (NEDHO) on the workings of the ATR NSUF. • Rory Kennedy, James Cole and Dan Ogden participated in a reactor instrumentation discussion with Jean-Francois Villard and Christopher Destouches of CEA and several members of the INL staff. • ATR NSUF received approval from the NE-20 office to start planning the annual Users Meeting. The meeting will be held at INL, June 22-25. • Mike Worley, director of the Office of Innovative Nuclear Research (NE-42), visited INL Nov. 4-5. Milestones Completed • Recommendations for the Summer Rapid Turnaround Experiment awards were submitted to DOE-HQ Nov. 12 (Level 2 milestone due Nov. 30). Major Accomplishments/Activities • The University of California, Santa Barbara 2 experiment was unloaded from the GE-2000 at HFEF. The experiment specimen packs will be removed and shipped to ORNL for PIE. • The Terrani experiment, one of three FY 2014 new awards, was completed utilizing the Advanced Photon Source MRCAT beamline. The experiment investigated the chemical state of Ag and Pd in SiC shell of irradiated TRISO particles via X-ray Absorption Fine Structure (XAFS) spectroscopy. Upcoming Meetings/Events • The ATR NSUF program review meeting will be held Dec. 9-10 at L’Enfant Plaza. In addition to NSUF staff and users, NE-4, NE-5 and NE-7 representatives will attend the meeting. Awarded Research Projects Boise State University Rapid Turnaround Experiments (14-485 and 14-486) Nanoindentation and TEM work on the T91, HT9, HCM12A and 9Cr ODS specimens has been completed at

  1. Experimental Study on Ash-Returned Reactor of CFB Atmospheric Air Gasifier

    Science.gov (United States)

    Shihong, Zhang; Luning, Tian; Xianrong, Zhou; Hanping, Chen; Haiping, Yang; Xianhua, Wang

    In an attempt to improve the gasification efficiency and decrease the carbon content in fly ash of atmospheric air CFB gasifiers, an innovatory equipment by name ash-returned reactor is put forward by SKLCC. Ash-returned reactor is an ash-returned apparatus on line of ash circulation, typically like "U" type valve in CFB boilers, with additional function of some extent combustion of residual carbon and increase the furnace inlet temperature of returning ash, and hence the coal conversion of gasifiers is enhanced. As to its configuration compared to conventional "U" type valve, ash-returned rector has two distinguished features of several times of height scale of fluidizing transportation region to meet the combustion reaction time need and appropriate heat transfer tube bundles arranged in the region to moderate the local temperature so as to avoid slagging. And hence, corresponding to the structure renovation, the material transportation and regulation performance of ash-returned reactor is primarily investigated through a series of experiments in a cold lab-scale facility in this paper. The heat transfer characteristic of the tube bundles is then researched and its influential factors are further discussed. These works lay a foundation on the following study of hot state experiments and industrial applications.

  2. Nuclear-fuel-cycle risk assessment: descriptions of representative non-reactor facilities. Sections 1-14

    Energy Technology Data Exchange (ETDEWEB)

    Schneider, K.J.

    1982-09-01

    The Fuel Cycle Risk Assessment Program was initiated to provide risk assessment methods for assistance in the regulatory process for nuclear fuel cycle facilities other than reactors. This report, the first from the program, defines and describes fuel cycle elements that are being considered in the program. One type of facility (and in some cases two) is described that is representative of each element of the fuel cycle. The descriptions are based on real industrial-scale facilities that are current state-of-the-art, or on conceptual facilities where none now exist. Each representative fuel cycle facility is assumed to be located on the appropriate one of four hypothetical but representative sites described. The fuel cycles considered are for Light Water Reactors with once-through flow of spent fuel, and with plutonium and uranium recycle. Representative facilities for the following fuel cycle elements are described for uranium (or uranium plus plutonium where appropriate): mining, milling, conversion, enrichment, fuel fabrication, mixed-oxide fuel refabrication, fuel reprocessing, spent fuel storage, high-level waste storage, transuranic waste storage, spent fuel and high-level and transuranic waste disposal, low-level and intermediate-level waste disposal, and transportation. For each representative facility the description includes: mainline process, effluent processing and waste management, facility and hardware description, safety-related information and potential alternative concepts for that fuel cycle element. The emphasis of the descriptive material is on safety-related information. This includes: operating and maintenance requirements, input/output of major materials, identification and inventories of hazardous materials (particularly radioactive materials), unit operations involved, potential accident driving forces, containment and shielding, and degree of hands-on operation.

  3. Experimental and kinetic modeling study of 3-methylheptane in a jet-stirred reactor

    KAUST Repository

    Karsenty, Florent

    2012-08-16

    Improving the combustion of conventional and alternative fuels in practical applications requires the fundamental understanding of large hydrocarbon combustion chemistry. The focus of the present study is on a high-molecular-weight branched alkane, namely, 3-methylheptane, oxidized in a jet-stirred reactor. This fuel, along with 2-methylheptane, 2,5-dimethylhexane, and n-octane, are candidate surrogate components for conventional diesel fuels derived from petroleum, synthetic Fischer-Tropsch diesel and jet fuels derived from coal, natural gas, and/or biomass, and renewable diesel and jet fuels derived from the thermochemical treatment of bioderived fats and oils. This study presents new experimental results along with a low- and high-temperature chemical kinetic model for the oxidation of 3-methylheptane. The proposed model is validated against these new experimental data from a jet-stirred reactor operated at 10 atm, over the temperature range of 530-1220 K, and for equivalence ratios of 0.5, 1, and 2. Significant effort is placed on the understanding of the effects of methyl substitution on important combustion properties, such as fuel reactivity and species formation. It was found that 3-methylheptane reacts more slowly than 2-methylheptane at both low and high temperatures in the jet-stirred reactor. © 2012 American Chemical Society.

  4. Experimental assessment of accident scenarios for the high temperature reactor fuel system

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O.; Avincola, V.; Bottomley, P.D.W.; Rondinella, V.V. [European Commission Joint Research Centre - Institute for Transuranium Elements (JRC-ITU) (Germany)

    2012-11-01

    The High Temperature Reactor (HTR) is an advanced reactor concept with particular safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRISO (tri-isotropic) coating designed to provide high fission product retention. Passive safety features of the HTR include a low power density in the core compared to other reactor designs; this ensures sufficient heat transport in a loss of coolant accident scenario. The temperature during such events would not exceed 1600 C, remaining well below the melting point of the fuel. An experimental assessment of the fuel behaviour under severe accident conditions is necessary to confirm the fission product retention of TRISO coated particles and to validate relevant computer codes. Though helium is used as coolant for the HTR system, additional corrosion effects come into play in case of an in-leakage affecting the primary circuit. The experimental scope of the present work focuses on two key aspects associated with the HTR fuel safety. Fission product retention at high temperatures (up to {proportional_to}1800 C) is analyzed with the so-called cold finger apparatus (KueFA: Kuehlfinger-Apparatur), while the performance of HTR fuel elements in case of air/steam ingress accidents is assessed with a high temperature corrosion apparatus (KORA: Korrosions-Apparatur). (orig.)

  5. Experimental measurement and CFD simulation on the hydrodynamics of an internal-loop airlift reactor

    Directory of Open Access Journals (Sweden)

    Liew Shi Yan

    2017-01-01

    Full Text Available This paper concerns with the experimental measurement and computational fluid dynamics simulation on local hydrodynamics of a gas-liquid internal-loop airlift reactor. The aim of this work is to study the sensitivity of the drag models and the significance of considering the lift force on the predictive accuracy of the simulation. The experimental analysis was carried out using laser Doppler anemometry at three different heights (i.e. Y = 0.20 m, 0.30 m and 0.38 m across the riser and downcomerat volumetric flow rate of 0.30 m3/h to provide validation for the simulation results. A transient three-dimensional gasliquid internal-loop airlift reactor was carried out using FLUENT 16.2 by implementing the two-fluid model approach. The Eulerian-Eulerian multiphase and standard κ-ε dispersed turbulence model wereemployed in this study. Results suggest that the spherical drag model performed poorly and that the drag model governed by Rayleigh-Taylor shows promising accuracy in the prediction of overall mean axial liquid velocity. On the other hand, the consideration of lift model shows slightly improvement in accuracy. These findings may serve as a guidance for future scale-up and design of airlift reactor studies

  6. Studies Related to the Oregon State University High Temperature Test Facility: Scaling, the Validation Matrix, and Similarities to the Modular High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schultz; Paul D. Bayless; Richard W. Johnson; William T. Taitano; James R. Wolf; Glenn E. McCreery

    2010-09-01

    The Oregon State University (OSU) High Temperature Test Facility (HTTF) is an integral experimental facility that will be constructed on the OSU campus in Corvallis, Oregon. The HTTF project was initiated, by the U.S. Nuclear Regulatory Commission (NRC), on September 5, 2008 as Task 4 of the 5 year High Temperature Gas Reactor Cooperative Agreement via NRC Contract 04-08-138. Until August, 2010, when a DOE contract was initiated to fund additional capabilities for the HTTF project, all of the funding support for the HTTF was provided by the NRC via their cooperative agreement. The U.S. Department of Energy (DOE) began their involvement with the HTTF project in late 2009 via the Next Generation Nuclear Plant project. Because the NRC interests in HTTF experiments were only centered on the depressurized conduction cooldown (DCC) scenario, NGNP involvement focused on expanding the experimental envelope of the HTTF to include steady-state operations and also the pressurized conduction cooldown (PCC). Since DOE has incorporated the HTTF as an ingredient in the NGNP thermal-fluids validation program, several important outcomes should be noted: 1. The reference prismatic reactor design, that serves as the basis for scaling the HTTF, became the modular high temperature gas-cooled reactor (MHTGR). The MHTGR has also been chosen as the reference design for all of the other NGNP thermal-fluid experiments. 2. The NGNP validation matrix is being planned using the same scaling strategy that has been implemented to design the HTTF, i.e., the hierarchical two-tiered scaling methodology developed by Zuber in 1991. Using this approach a preliminary validation matrix has been designed that integrates the HTTF experiments with the other experiments planned for the NGNP thermal-fluids verification and validation project. 3. Initial analyses showed that the inherent power capability of the OSU infrastructure, which only allowed a total operational facility power capability of 0.6 MW, is

  7. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  8. The experimental facility of Tournemire; La station experimentale de Tournemire

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-10-01

    This document presents the underground facility of Tournemire (Aveyron, France). The Tournemire abandoned railway tunnel gives access to a 250 m thick Jurassic clay bed covered with 250 m of limestones. The main goal of the Tournemire project is the study of the mechanical properties and fracturing of a clay formation and of its ability to be used as a deep underground storage facility for radioactive wastes. The document comprises a general presentation brochure and a description of the geologic, tectonic, geomechanical and hydro-geochemical surveys carried out in the facility. (J.S.)

  9. Experimental Geothermal Research Facilities study (Phase 0). Final report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1974-12-31

    Phase 0 of a project for Experimental Geothermal Research Facilities focuses on identification of a representative liquid-dominated geothermal reservoir of moderate temperature and salinity, preliminary engineering design of an appropriate energy conversion system, identification of critical technology, and planning for implementation of experimental facilities. The objectives included development of liaison with the industrial sector, to ensure responsiveness to their views in facility requirements and planning, and incorporation of environmental and socioeconomic factors. The overall project, of which the six-month Phase 0 is reported herein, is phased in accordance with RANN guidelines. This Phase 0 report covers problem definition and systems requirements. Phase 1 will involve design of the experimental facility, and testing of components. Phase 2 will comprise detailed design and construction of an experimental geothermal electrical powerplant at East Mesa, Imperial County, California. Facilities will incorporate capability for research in component, system, and materials technology and a nominal 10 MWe experimental, binary cycle, power generating plant.

  10. Experimental conditions for determination of the neutrino mass hierarchy with reactor antineutrinos

    Directory of Open Access Journals (Sweden)

    Myoung Youl Pac

    2016-01-01

    Full Text Available This article reports the optimized experimental requirements to determine neutrino mass hierarchy using electron antineutrinos (ν¯e generated in a nuclear reactor. The features of the neutrino mass hierarchy can be extracted from the |Δm312| and |Δm322| oscillations by applying the Fourier sine and cosine transforms to the L/E spectrum. To determine the neutrino mass hierarchy above 90% probability, the requirements on the energy resolution as a function of the baseline are studied at sin2⁡2θ13=0.1. If the energy resolution of the neutrino detector is less than 0.04/Eν and the determination probability obtained from Bayes' theorem is above 90%, the detector needs to be located around 48–53 km from the reactor(s to measure the energy spectrum of ν¯e. These results will be helpful for setting up an experiment to determine the neutrino mass hierarchy, which is an important problem in neutrino physics.

  11. Development of Experimental Facilities for Advanced Spent Fuel Management Technology

    Energy Technology Data Exchange (ETDEWEB)

    You, G. S.; Jung, W. M.; Ku, J. H. [and others

    2004-07-01

    The advanced spent fuel management process(ACP), proposed to reduce the overall volume of the PWR spent fuel and improve safety and economy of the long-term storage of spent fuel, is under research and development. This technology convert spent fuels into pure metal-base uranium with removing the highly heat generating materials(Cs, Sr) efficiently and reducing of the decay heat, volume, and radioactivity from spent fuel by 1/4. In the next phase(2004{approx}2006), the demonstration of this technology will be carried out for verification of the ACP in a laboratory scale. For this demonstration, the hot cell facilities of {alpha}-{gamma} type and auxiliary facilities are required essentially for safe handling of high radioactive materials. As the hot cell facilities for demonstration of the ACP, a existing hot cell of {beta}-{gamma} type will be refurbished to minimize construction expenditures of hot cell facility. In this study, the design requirements are established, and the process detail work flow was analysed for the optimum arrangement to ensure effective process operation in hot cell. And also, the basic and detail design of hot cell facility and process, and safety analysis was performed to secure conservative safety of hot cell facility and process.

  12. ESTIMATION OF ROUTINE DISCHARGE OF RADIONUCLIDES ON POWER REACTOR EXPERIMENTAL RDE

    Directory of Open Access Journals (Sweden)

    P.M. Udiyani

    2017-02-01

    Full Text Available Experimental power reactor (RDE which is planned to be constructed by BATAN is a kind of High Temperature Gas Cooled Reactor (HTGR with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products remained in the core. Although the fission products released into the environment are very small, in order to comply the regulations the study about environmental radiation on normal or routine operation condition need to be performed. Estimation of radiology in the environment involves the source term released into the environment under routine operation condition. The purpose of this study is to estimate the source term released into the environment based on postulation of normal or routine operations of RDE. The research approach starts with an assumption that there are defects and impurities in the TRISO fuel because of limitation during the fabrication. Mechanism of fission products release from the fuel to the environment was created based on the safety features design of RDE. Radionuclides inventories in the reactor were calculated using ORIGEN-2 whose library has been modified for HTGR type, and the assumptions of defects of the TRISO fuel and release fraction for each compartment of RDE safety system used a reference parameter. The results showed that the important source terms of RDE are group of noble gases (Kr and Xe, halogen (I, Sr, Cs, H-3, and Ag. Activities of RDE source terms for routine operations have no significant difference with the HTGR source terms with the same power. Keywords: routine discharge, radionuclide, source term, RDE, HTGR

  13. Modeling and Experimental Studies of Mercury Oxidation and Adsorption in a Fixed-Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Buitrago, Paula A.; Morrill, Mike; Lighty, JoAnn S.; Silcox, Geoffrey D.

    2009-06-15

    This report presents experimental and modeling mercury oxidation and adsorption data. Fixed-bed and single-particle models of mercury adsorption were developed. The experimental data were obtained with two reactors: a 300-W, methane-fired, tubular, quartz-lined reactor for studying homogeneous oxidation reactions and a fixed-bed reactor, also of quartz, for studying heterogeneous reactions. The latter was attached to the exit of the former to provide realistic combustion gases. The fixed-bed reactor contained one gram of coconut-shell carbon and remained at a temperature of 150°C. All methane, air, SO2, and halogen species were introduced through the burner to produce a radical pool representative of real combustion systems. A Tekran 2537A Analyzer coupled with a wet conditioning system provided speciated mercury concentrations. At 150°C and in the absence of HCl or HBr, the mercury uptake was about 20%. The addition of 50 ppm HCl caused complete capture of all elemental and oxidized mercury species. In the absence of halogens, SO2 increased the mercury adsorption efficiency to up to 30 percent. The extent of adsorption decreased with increasing SO2 concentration when halogens were present. Increasing the HCl concentration to 100 ppm lessened the effect of SO2. The fixed-bed model incorporates Langmuir adsorption kinetics and was developed to predict adsorption of elemental mercury and the effect of multiple flue gas components. This model neglects intraparticle diffusional resistances and is only applicable to pulverized carbon sorbents. It roughly describes experimental data from the literature. The current version includes the ability to account for competitive adsorption between mercury, SO2, and NO2. The single particle model simulates in-flight sorbent capture of elemental mercury. This model was developed to include Langmuir and Freundlich isotherms, rate equations, sorbent feed rate, and

  14. Overview of the TIBER 2 (Thermal Ignition/Burn Experimental Reactor) design

    Science.gov (United States)

    Henning, C. D.; Logan, B. G.

    1987-10-01

    The TIBER 2 Tokamak Ignition/Burn Experimental Reactor design is the result of efforts by numerous people and institutions, including many fusion laboratories, universities, and industries. While subsystems will be covered extensively in other reports, this overview will attempt to place the work in perspective. Major features of the design are compact size, low cost, and steady-state operation. These are achieved through plasma shaping and innovative features such as radiation tolerant magnets and optimized shielding. While TIBER 2 can operate in a pulsed mode, steady-state is preferred for nuclear testing. Current drive is achieved by a combination of lower hybrid and neutral beams. In addition, 10 MW of ECR is added for disruption control and current drive profiling. The TIBER 2 design has been the US option in preparation for the International Thermonuclear Experimental Reactor (ITER). Other equivalent national designs are the NET in Europe, the FER in Japan and the OTR in the USSR. These designs will help set the basis for the new international design effort.

  15. Modeling and experimental validation of hydrodynamics in an ultrasonic batch reactor.

    Science.gov (United States)

    Ajmal, M; Rusli, S; Fieg, G

    2016-01-01

    Simulation of hydrodynamics in ultrasonic batch reactor containing immobilized enzymes as catalyst is done. A transducer with variable power and constant frequency (24 kHz) is taken as source of ultrasound (US). Simulation comprises two steps. In first step, acoustic pressure field is simulated and in second step effect of this field on particle trajectories is simulated. Simulation results are compared with experimentally determined particle trajectories using PIV Lab (particle image velocimetry). Effect of varying ultrasonic power, positioning and number of ultrasonic sources on particle trajectories is studied. It is observed that catalyst particles tend to orientate according to pattern of acoustic pressure field. An increase in ultrasonic power increases particle velocity and also brings more particles into motion. Simulation results are found to be in agreement with experimentally determined data. Copyright © 2015 Elsevier B.V. All rights reserved.

  16. Kinetics of vinyl acetate emulsion polymerization in a pulsed tubular reactor: comparison between experimental and simulation results

    Directory of Open Access Journals (Sweden)

    Sayer C.

    2002-01-01

    Full Text Available A new reactor, the pulsed sieve plate column (PSPC, was developed to perform continuous emulsion polymerization reactions. This reactor combines the enhanced flexibility of tubular reactors with the mixing behavior provided by sieved plates and by the introduction of pulses that is important to prevent emulsion destabilization. The main objective of this work is to study the kinetics of vinyl acetate (VA emulsion polymerization reactions performed in this PSPC. Therefore, both experimental studies and reaction simulations were performed. Results showed that it is possible to obtain high conversions with rather low residence times in the PSPC.

  17. CHARACTERIZATION OF A PRECIPITATE REACTOR FEED TANK (PRFT) SAMPLE FROM THE DEFENSE WASTE PROCESSING FACILITY (DWPF)

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C.; Bannochie, C.

    2014-05-12

    A sample of from the Defense Waste Processing Facility (DWPF) Precipitate Reactor Feed Tank (PRFT) was pulled and sent to the Savannah River National Laboratory (SRNL) in June of 2013. The PRFT in DWPF receives Actinide Removal Process (ARP)/ Monosodium Titanate (MST) material from the 512-S Facility via the 511-S Facility. This 2.2 L sample was to be used in small-scale DWPF chemical process cell testing in the Shielded Cells Facility of SRNL. A 1L sub-sample portion was characterized to determine the physical properties such as weight percent solids, density, particle size distribution and crystalline phase identification. Further chemical analysis of the PRFT filtrate and dissolved slurry included metals and anions as well as carbon and base analysis. This technical report describes the characterization and analysis of the PRFT sample from DWPF. At SRNL, the 2.2 L PRFT sample was composited from eleven separate samples received from DWPF. The visible solids were observed to be relatively quick settling which allowed for the rinsing of the original shipping vials with PRFT supernate on the same day as compositing. Most analyses were performed in triplicate except for particle size distribution (PSD), X-ray diffraction (XRD), Scanning Electron Microscopy (SEM) and thermogravimetric analysis (TGA). PRFT slurry samples were dissolved using a mixed HNO3/HF acid for subsequent Inductively Coupled Plasma Atomic Emission Spectroscopy (ICPAES) and Inductively Coupled Plasma Mass Spectroscopy (ICP-MS) analyses performed by SRNL Analytical Development (AD). Per the task request for this work, analysis of the PRFT slurry and filtrate for metals, anions, carbon and base were primarily performed to support the planned chemical process cell testing and to provide additional component concentrations in addition to the limited data available from DWPF. Analysis of the insoluble solids portion of the PRFT slurry was aimed at detailed characterization of these solids (TGA, PSD

  18. Superconducting magnet system for an experimental disk MHD facility

    NARCIS (Netherlands)

    Knoopers, H.G.; ten Kate, Herman H.J.; van de Klundert, L.J.M.; van de Klundert, L.J.M.

    1991-01-01

    A predesign of a split-pair magnet for a magnetohydrodynamic (MHD) facility for testing a 10-MW open-cycle disk or a 5-MW closed-cycle disk generator is presented. The magnet system consists of a NbTi and a Nb 3Sn section, which provide a magnetic field of 9 T in the active area of the MHD channel.

  19. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    Nigg, D.W.; Schwint, A.E.; Hartwell, J.K.; Heber, E.M.; Trivillin, V.; Castillo, J.; Wentzeis, L.; Sloan, P.; Wemple, C.A.

    2004-10-04

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  20. Collaborative Physical and Biological Dosimetry Studies for Neutron Capture Therapy at the RA-1 Research Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Amanda E. Schwint; John K. Hartwell; Elisa M. Heber; Veronica Trivillin; Jorge Castillo; Luis Wentzeis; Patrick Sloan; Charles A. Wemple

    2004-10-01

    Initial physical dosimetry measurements have been completed using activation spectrometry and thermoluminiscent dosimeters to characterize the BNCT irradiation facility developed at the RA-1 research reactor operated by the Argentine National Atomic Energy Commission in Buenos Aires. Some biological scoping irradiations have also been completed using a small-animal (hamster) oral mucosa tumor model. Results indicate that the RA-1 neutron source produces useful dose rates but that some improvements in the initial configuration will be needed to optimize the spectrum for thermal-neutron BNCT research applications.

  1. Experimental study and modeling of a high-temperature solar chemical reactor for hydrogen production from methane cracking

    Energy Technology Data Exchange (ETDEWEB)

    Abanades, Stephane; Flamant, Gilles [Processes, Materials, and Solar Energy Laboratory, CNRS (PROMES-CNRS, UPR 8521), 7 Rue du Four Solaire, 66120 Odeillo Font-Romeu (France)

    2007-07-15

    A high-temperature fluid-wall solar reactor was developed for the production of hydrogen from methane cracking. This laboratory-scale reactor features a graphite tubular cavity directly heated by concentrated solar energy, in which the reactive flowing gas dissociates to form hydrogen and carbon black. The solar reactor characterization was achieved with: (a) a thorough experimental study on the reactor performance versus operating conditions and (b) solar reactor modeling. The results showed that the conversion of CH{sub 4} and yield of H{sub 2} can exceed 97% and 90%, respectively, and these depend strongly on temperature and on fluid-wall heat transfer and reaction surface area. In addition to the experimental study, a 2D computational model coupling transport phenomena was developed to predict the mapping of reactor temperature and of species concentration, and the reaction extent at the outlet. The model was validated and kinetics of methane decomposition were identified from simulations and comparison to experimental results. (author)

  2. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

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

  3. NWFSC OA facility water chemistry - Ocean acidification species exposure experimental facility

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — We have developed a unique facility for conducting high-quality experiments on marine organisms in seawater with controlled carbon chemistry conditions. The...

  4. Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Nuclear Reactor Lab.)

    1989-09-01

    This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power on spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.

  5. Radiological survey support activities for the decommissioning of the Ames Laboratory Research Reactor Facility, Ames, Iowa

    Energy Technology Data Exchange (ETDEWEB)

    Wynveen, R.A.; Smith, W.H.; Sholeen, C.M.; Justus, A.L.; Flynn, K.F.

    1984-09-01

    At the request of the Engineering Support Division of the US Department of Energy-Chicago Operations Office and in accordance with the programmatic overview/certification responsibilities of the Department of Energy Environmental and Safety Engineering Division, the Argonne National Laboratory Radiological Survey Group conducted a series of radiological measurements and tests at the Ames Laboratory Research Reactor located in Ames, Iowa. These measurements and tests were conducted during 1980 and 1981 while the reactor building was being decontaminated and decommissioned for the purpose of returning the building to general use. The results of these evaluations are included in this report. Although the surface contamination within the reactor building could presumably be reduced to negligible levels, the potential for airborne contamination from tritiated water vapor remains. This vapor emmanates from contamination within the concrete of the building and should be monitored until such time as it is reduced to background levels. 2 references, 8 figures, 6 tables.

  6. PERGOLA: an experimental facility to investigate storable propellants' combustion

    OpenAIRE

    Boust, Bastien; Bellenoue, Marc; Lamory, Stéphane; Labarthe, Emilie; Girard, Nathalie

    2017-01-01

    International audience; PPRIME Institute and CNES have initiated research activities to investigate the propulsion behaviour of liquid storable propellants, and the related phenomena: atomization of propellants, ignition ability, combustion stability, propulsion efficiency. For this purpose, a specific facility " PERGOLA " has been designed with high modularity, allowing physical diagnostics in a pressurized combustor featuring relevant flow conditions (a few hundred g/s, 5 MPa). The first re...

  7. First experimental data of the cryogenic safety test facility PICARD

    Science.gov (United States)

    Heidt, C.; Henriques, A.; Stamm, M.; Grohmann, S.

    2017-02-01

    The test facility PICARD, which stands for Pressure Increase in Cryostats and Analysis of Relief Devices, has been designed and constructed for cryogenic safety experiments. With a cryogenic liquid volume of 100 L, a nominal design pressure of 16 bar(g) and the capacity of measuring helium mass flow rates through safety relief devices up to 4 kg/s, the test facility allows the systematic investigation of hazardous incidents in cryostats under realistic conditions. In the course of experiments, the insulating vacuum is vented with atmospheric air or gaseous nitrogen at ambient temperature under variation of the venting diameter, the thermal insulation, the cryogenic fluid, the liquid level and the set pressure in order to analyze the impact on the heat flux and hence on the process dynamics. A special focus will be on the occurrence and implications of two-phase flow during expansion and on measuring the flow coefficients of safety devices at cryogenic temperatures. This paper describes the commissioning and the general performance of the test facility at liquid helium temperatures. Furthermore, the results of first venting experiments are presented.

  8. Large Object Irradiation Facility In The Tangential Channel Of The JSI TRIGA Reactor

    CERN Document Server

    Radulovic, Vladimir; Kaiba, Tanja; Kavsek, Darko; Cindro, Vladimir; Mikuz, Marko; Snoj, Luka

    2017-01-01

    This paper presents the design and installation of a new irradiation device in the Tangential Channel of the JSI TRIGA reactor in Ljubljana, Slovenia. The purpose of the device is to enable on-line irradiation testing of electronic components considerably larger in size (of lateral dimensions of at least 12 cm) than currently possible in the irradiation channels located in the reactor core, in a relatively high neutron flux (exceeding 10^12 n cm^-2 s^-1) and to provide adequate neutron and gamma radiation shielding.

  9. Advanced automation concepts applied to Experimental Breeder Reactor-II startup

    Energy Technology Data Exchange (ETDEWEB)

    Berkan, R.C.; Upadhyaya, B.R.; Bywater, R.L. (Tennessee Univ., Knoxville, TN (United States). Dept. of Nuclear Engineering); Kisner, R.A. (Oak Ridge National Lab., TN (United States))

    1991-08-01

    The major objective of this work is to demonstrate through simulations that advanced liquid-metal reactor plants can be operated from low power by computer control. Development of an automatic control system with this objective will help resolve specific issues and provide proof through demonstration that automatic control for plant startup is feasible. This paper presents an advanced control system design for startup of the Experimental Breeder Reactor-2 (EBR-2) located at Idaho Falls, Idaho. The design incorporates recent methods in nonlinear control with advanced diagnostics techniques such as neural networks to form an integrated architecture. The preliminary evaluations are obtained in a simulated environment by a low-order, valid nonlinear model. Within the framework of phase 1 research, the design includes an inverse dynamics controller, a fuzzy controller, and an artificial neural network controller. These three nonlinear control modules are designed to follow the EBR-2 startup trajectories in a multi-input/output regime. They are coordinated by a supervisory routine to yield a fault-tolerant, parallel operation. The control system operates in three modes: manual, semiautomatic, and fully automatic control. The simulation results of the EBR-2 startup transients proved the effectiveness of the advanced concepts. The work presented in this paper is a preliminary feasibility analysis and does not constitute a final design of an automated startup control system for EBR-2. 14 refs., 43 figs.

  10. Measurements Methods for the analysis of Nuclear Reactors Thermal Hydraulic in Water Scaled Facilities

    Science.gov (United States)

    Spaccapaniccia, C.; Planquart, P.; Buchlin, J. M. AB(; ), AC(; )

    2018-01-01

    The Belgian nuclear research institute (SCK•CEN) is developing MYRRHA. MYRRHA is a flexible fast spectrum research reactor, conceived as an accelerator driven system (ADS). The configuration of the primary loop is pool-type: the primary coolant and all the primary system components (core and heat exchangers) are contained within the reactor vessel, while the secondary fluid is circulating in the heat exchangers. The primary coolant is Lead Bismuth Eutectic (LBE). The recent nuclear accident of Fukushima in 2011 changed the requirements for the design of new reactors, which should include the possibility to remove the residual decay heat through passive primary and secondary systems, i.e. natural convection (NC). After the reactor shut down, in the unlucky event of propeller failures, the primary and secondary loops should be able to remove the decay heat in passive way (Natural Convection). The present study analyses the flow and the temperature distribution in the upper plenum by applying laser imaging techniques in a laboratory scaled water model. A parametric study is proposed to study stratification mitigation strategies by varying the geometry of the buffer tank simulating the upper plenum.

  11. Experimental monitoring of ozone production in a PET cyclotron facility

    Energy Technology Data Exchange (ETDEWEB)

    Zanibellato, L. [Department of Energetic, Nuclear and Environmental Control, University of Bologna (Italy); Cicoria, G.; Pancaldi, D. [Department of Medical Physics, University Hospital ' S.Orsola-Malpighi' , Bologna (Italy); Boschi, S. [PET Radiopharmacy Unit, Department of Nuclear Medicine, University Hospital ' S.Orsola-Malpighi' , Bologna (Italy); Mostacci, D. [Department of Energetic, Nuclear and Environmental Control, University of Bologna (Italy); Marengo, M., E-mail: marengo@med.unibo.i [Department of Medical Physics, University Hospital ' S.Orsola-Malpighi' , Bologna (Italy)

    2010-10-15

    Ozone produced from radiolytic processes was investigated as a possible health hazard in the working environment at the University Hospital 'S.Orsola-Malpighi' PET facility. Intense radiation fields can generate ozone, known to be the most toxic gas produced by ionizing radiation around a particle accelerator. To evaluate ozone concentration in air, two different measurement campaigns were conducted with passive diffusion detectors. Comparison of the results with the concentration limits recommended by American Conference of Governmental Industrial Hygienists (ACGIH) demonstrated that ozone poses no health hazard to workers around a biomedical cyclotron.

  12. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1992; Twenty-fifth annual report, Volume 14

    Energy Technology Data Exchange (ETDEWEB)

    Raddatz, C.T. [US Nuclear Regulatory Commission, Washington, DC (United States). Division of Regulatory Applications; Hagemeyer, D. [Science Applications International Corp., Oak Ridge, TN (United States)

    1993-12-01

    This report summarizes the occupational radiation exposure information that has been reported to the NRC`s Radiation Exposure Information Reporting System (REIRS) by nuclear power facilities and certain other categories of NRC licensees during the years 1969 through 1992. The bulk of the data presented in the report was obtained from annual radiation exposure reports submitted in accordance with the requirements of 10CFR20.407 and the technical specifications of nuclear power plants. Data on workers terminating their employment at certain NRC licensed facilities were obtained from reports submitted pursuant to 10CFR20.408. The 1992 annual reports submitted by about 364 licensees indicated that approximately 204,365 individuals were monitored, 183,927 of whom were monitored by nuclear power facilities. They incurred an average individual dose of 0.16 rem (cSv) and an average measurable dose of about 0.30 (cSv). Termination radiation exposure reports were analyzed to reveal that about 74,566 individuals completed their employment with one or more of the 364 covered licensees during 1992. Some 71,846 of these individuals terminated from power reactor facilities, and about 9,724 of them were considered to be transient workers who received an average dose of 0.50 rem (cSv).

  13. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    coolants. The purpose of the high temperature helium loop (HTHL) is to simulate technical and chemical conditions of VHTR's coolant. The loop is intended to serve an as experimental device for fatigue and creep tests of construction metallic materials for gas-cooled reactors and it should be also employed for research in field of gaseous coolant chemistry. The loop will serve also for tests of nuclear graphite, dosing and helium purification systems. Because the VHTR is a new reactor concept, major technical uncertainties remain relative to helium-cooled advanced reactor systems. This paper summarizes also the concept of the HTHL in the Research Centre Rez Ltd., its design, utilization and future plans for experimental setup.

  14. Conceptual design of superconducting magnet systems for the Argonne Tokamak Experimental Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, S.T.; Turner, L.R.; Mills, F.E.; DeMichele, D.W.; Smelser, P.; Kim, S.H.

    1976-01-01

    As an integral effort in the Argonne Tokamak Experimental Power Reactor Conceptual Design, the conceptual design of a 10-tesla, pure-tension superconducting toroidal-field (TF) coil system has been developed in sufficient detail to define a realistic design for the TF coil system that could be built based upon the current state of technology with minimum technological extrapolations. A conceptual design study on the superconducting ohmic-heating (OH) coils and the superconducting equilibrium-field (EF) coils were also completed. These conceptual designs are developed in sufficient detail with clear information on high current ac conductor design, cooling, venting provision, coil structural support and zero loss poloidal coil cryostat design. Also investigated is the EF penetration into the blanket and shield.

  15. A large scale fullerenes synthesis solar reactor modelling and first experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Guillard, T.; Flamand, G.; Robert, J.F.; Rivoire, B.; Olalde, G.; Alvarez, L. [Centre National de la Recherche Scientifique (CNRS-IMP), 66 - Font-Romeu (France); Laplaze, D. [Universite de Montpellier, GDPC, 34 (France)

    1999-03-01

    After the promising results obtained with a 2 kW solar furnace for fullerenes and nano-tubes synthesis, a large scale production project using the 1 MW Odeillo solar furnace started in 1997. This paper presents the first experimental results obtained with a concept-validation vessel and the comparison with a numerical simulation of the target thermal behavior. It is shown that a 6 mm i.d. graphite rod heated by a 500 W/cm{sup 2} incident solar flux density (I{sub s}) reaches a front temperature of 2800 K, in agreement with the thermal model. On this basis, accurate prediction of maximum working temperature of the 1 MW reactor is proposed: 3400 K for I{sub s} = 900 W/cm{sup 2}. (authors)

  16. Review of the International Thermonuclear Experimental Reactor (ITER) detailed design report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-18

    Dr. Martha Krebs, Director, Office of Energy Research at the US Department of Energy (DOE), wrote to the Fusion Energy Sciences Advisory Committee (FESAC), in letters dated September 23 and November 6, 1996, requesting that FESAC review the International Thermonuclear Experimental Reactor (ITER) Detailed Design Report (DDR) and provide its view of the adequacy of the DDR as part of the basis for the United States decision to enter negotiations with the other interested Parties regarding the terms and conditions for an agreement for the construction, operations, exploitation and decommissioning of ITER. The letter from Dr. Krebs, referred to as the Charge Letter, provided context for the review and a set of questions of specific interest.

  17. Development plan for the External Hazards Experimental Group. Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, Justin Leigh [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis Lee [Idaho National Lab. (INL), Idaho Falls, ID (United States); Burns, Douglas Edward [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kammerer, Annie [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This report describes the development plan for a new multi-partner External Hazards Experimental Group (EHEG) coordinated by Idaho National Laboratory (INL) within the Risk-Informed Safety Margin Characterization (RISMC) technical pathway of the Light Water Reactor Sustainability Program. Currently, there is limited data available for development and validation of the tools and methods being developed in the RISMC Toolkit. The EHEG is being developed to obtain high-quality, small- and large-scale experimental data validation of RISMC tools and methods in a timely and cost-effective way. The group of universities and national laboratories that will eventually form the EHEG (which is ultimately expected to include both the initial participants and other universities and national laboratories that have been identified) have the expertise and experimental capabilities needed to both obtain and compile existing data archives and perform additional seismic and flooding experiments. The data developed by EHEG will be stored in databases for use within RISMC. These databases will be used to validate the advanced external hazard tools and methods.

  18. A new MTR fuel for a new MTR reactor: UMo for the Jules Horowitz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guigon, B. [CEA Cadarache, Dir. de l' Energie Nucleaire DEN, Reacteur Jules Horowitz, 13 - Saint-Paul-lez-Durance (France); Vacelet, H. [Compagnie pour l' Etude et la Realisation de Combustibles Atomiques, CERCA, Etablissement de Romans, 26 (France); Dornbusch, D. [Technicatome, Service d' Architecture Generale, 13 - Aix-en-Provence (France)

    2003-07-01

    Within some years, the Jules Horowitz Reactor will be the only working experimental reactor (material and fuel testing reactor) in France. It will have to provide facilities for a wide range of needs: from activation analysis to power reactor fuel qualification. In this paper will be presented the main characteristics of the Jules Horowitz Reactor: its total power, neutron flux, fuel element... Safety criteria will be explained. Finally merits and disadvantages of UMo compared to the standard U{sub 3}Si{sub 2} fuel will be discussed. (authors)

  19. Participation in the U.S. Department of Energy Reactor Sharing Program. Progress report

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed here.

  20. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

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

  1. Computational analysis of the dose rates at JSI TRIGA reactor irradiation facilities.

    Science.gov (United States)

    Ambrožič, K; Žerovnik, G; Snoj, L

    2017-12-01

    The JSI TRIGA Mark II, IJS research reactor is equipped with numerous irradiation positions, where samples can be irradiated by neutrons and γ-rays. Irradiation position selection is based on its properties, such as physical size and accessibility, as well as neutron and γ-ray spectra, flux and dose intensities. This paper presents an overview on the neutron and γ-ray fluxes, spectra and dose intensities calculations using Monte Carlo MCNP software and ENDF/B-VII.0 nuclear data libraries. The dose-rates are presented in terms of ambient dose equivalents, air kerma, and silicon dose equivalent. At full reactor power the neutron ambient dose equivalent ranges from 5.5×103Svh-1 to 6×106Svh-1, silicon dose equivalent from 6×102Gy/hsi to 3×105Gy/hsi, and neutron air kerma from 4.3×103Gyh-1 to 2×105Gyh-1. Ratio of fast (1MeVreactor power from 3.4×103Svh-1 to 3.6×105Svh-1 and γ air kerma range 3.1×103Gyh-1 to 2.9×105Gyh-1. Copyright © 2017 Elsevier Ltd. All rights reserved.

  2. On the possible use of the MASURCA reactor as a flexible, high-intensity, fast neutron beam facility

    Science.gov (United States)

    Dioni, Luca; Jacqmin, Robert; Sumini, Marco; Stout, Brian

    2017-09-01

    In recent work [1, 2], we have shown that the MASURCA research reactor could be used to deliver a fairly-intense continuous fast neutron beam to an experimental room located next to the reactor core. As a consequence of the MASURCA favorable characteristics and diverse material inventories, the neutron beam intensity and spectrum can be further tailored to meet the users' needs, which could be of interest for several applications. Monte Carlo simulations have been performed to characterize in detail the extracted neutron (and photon) beam entering the experimental room. These numerical simulations were done for two different bare cores: A uranium metallic core (˜30% 235U enriched) and a plutonium oxide core (˜25% Pu fraction, ˜78% 239Pu). The results show that the distinctive resonance energy structures of the two core leakage spectra are preserved at the channel exit. As the experimental room is large enough to house a dedicated set of neutron spectrometry instruments, we have investigated several candidate neutron spectrum measurement techniques, which could be implemented to guarantee well-defined, repeatable beam conditions to users. Our investigation also includes considerations regarding the gamma rays in the beams.

  3. Experimental underground facility to evaluate remote sensing instruments

    Science.gov (United States)

    Tatry, Veronique; Bocquet, Nathalie; Menard, Tamara; Nomine, Michel

    1999-09-01

    To carry out metrological tests on the automatic devices used in air quality monitoring networks, INERIS follows a standard text. In order to apply it, the instruments have been fed with reference mixtures of gases in several reproducible concentrations. Compliance with this requirement is quite easy to obtain, in the case of conventional techniques, but the problems remain with optical remote sensing instruments. The difficulties can be solved by using the Beer-Lambert law indicating that the product of the concentration by the distance is a constant value. To implement this optical law, INERIS uses a system with a two meters glass cell with quartz windows inserted in the instrument beam and fed dynamically with known concentrations of reference mixtures of gases. The test-bench facility is set up in a tunnel 90 meters long in INERIS. Tests on SO2, NO2 and O3 were carried out in 1997 on two types of DOAS systems set up in parallel: an OPSIS system and an Environment SA SANOA system. The following features were tested: the detection limit, the linearity and the drift. The tests results on the response of the two types of instruments are in agreement with the results obtained by the ERLAP at the Joint Research Center in Ispra.

  4. Structures and Materials Experimental Facilities and Capabilities Catalog

    Science.gov (United States)

    Horta, Lucas G. (Compiler); Kurtz-Husch, Jeanette D. (Compiler)

    2000-01-01

    The NASA Center of Excellent for Structures and Materials at Langley Research Center is responsible for conducting research and developing useable technology in the areas of advanced materials and processing technologies, durability, damage tolerance, structural concepts, advanced sensors, intelligent systems, aircraft ground operations, reliability, prediction tools, performance validation, aeroelastic response, and structural dynamics behavior for aerospace vehicles. Supporting the research activities is a complementary set of facilities and capabilities documented in this report. Because of the volume of information, the information collected was restricted in most cases to one page. Specific questions from potential customers or partners should be directed to the points of contacts provided with the various capabilities. Grouping of the equipment is by location as opposed to function. Geographical information of the various buildings housing the equipment is also provided. Since this is the first time that such an inventory is ever collected at Langley it is by no means complete. It is estimated that over 90 percent of the equipment capabilities at hand are included but equipment is continuously being updated and will be reported in the future.

  5. Integration of experimental facilities: A joint effort for establishing a common knowledge base in experimental work on hydrogen safety

    NARCIS (Netherlands)

    Reinecke, E.A.; Huebert, T.; Tkatschenko, I.; Kessler, A.; Kuznetsov, M.; Wilkins, M.; Hedley, D.; Azkarate, I.; Proust, C.; Acosta-Iborra, B.; Gavrikov, B.; Bruijn, P.C.J. de; Marangon, A.; Teodorczyk, A.; Grafwallner, A.

    2011-01-01

    In the area of hydrogen safety, research facilities are essential for the experimental investigation of relevant phenomena, for testing devices and safety concepts, as well as for the generation of validation data for the various numerical codes and models. Within the framework of the European

  6. Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

    Science.gov (United States)

    Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

    2017-07-01

    Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

  7. Analysis of the optimal fuel composition for the Indonesian experimental power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liem, Peng Hong [Nippon Advanced Information Service (NAIS Co., Inc.), Ibaraki (Japan); Sembiring, Tagor Malem [National Nuclear Energy Agency of Indonesia, Banten (Indonesia). Center for Nuclear Reactor Technology and Safety; Arbie, Bakri; Subki, Iyos [PT MOTAB Technology, Jakarta Barat (Indonesia)

    2017-03-15

    The optimal fuel composition of the 10 MWth Experimental Power Reactor (RDE), to be built by the Indonesian National Nuclear Energy Agency (BATAN), is a very important design parameter since it will directly affect the fuel cost, new and spent fuel storage capacity, and other back-end environmental burden. The RDE is a very small sized pebble-bed high temperature gas-cooled reactor (HTGR) with low enriched uranium (LEU) UO{sub 2} TRISO fuel under multipass or once-through-then-out fueling scheme. A scoping study on fuel composition parameters, namely heavy metal (HM) loading per pebble and uranium enrichment is conducted. All burnup, criticality calculations and core equilibrium search are carried out by using BATAN-MPASS, a general in-core fuel management code for pebble bed HTGRs, featured with many automatic equilibrium searching options as well as thermal-hydraulic calculation capability. The RDE User Requirement Document issued by BATAN is used to derive the main core design parameters and constraints. The scoping study is conducted over uranium enrichment in the range of 10 to 20 w/o and HM loading in the range of 4 g to 10 g/pebble. Fissile loading per unit energy generated (kg/GWd) is taken as the objective function for the present scoping study. The analysis results show that the optimal HM loading is around 8 g/pebble. Under the constraint of 80 GWd/t fuel discharge burnup imposed by the technical specification, the uranium enrichment for the optimal HM loading is approximately 13 w/o.

  8. Thermo-kinetic instabilities in model reactors. Examples in experimental tests

    Science.gov (United States)

    Lavadera, Marco Lubrano; Sorrentino, Giancarlo; Sabia, Pino; de Joannon, Mara; Cavaliere, Antonio; Ragucci, Raffaele

    2017-11-01

    The use of advanced combustion technologies (such as MILD, LTC, etc.) is among the most promising methods to reduce emission of pollutants. For such technologies, working temperatures are enough low to boost the formation of several classes of pollutants, such as NOx and soot. To access this temperature range, a significant dilution as well as preheating of reactants is required. Such conditions are usually achieved by a strong recirculation of exhaust gases that simultaneously dilute and pre-heat the fresh reactants. These peculiar operative conditions also imply strong fuel flexibility, thus allowing the use of low calorific value (LCV) energy carriers with high efficiency. However, the intersection of low combustion temperatures and highly diluted mixtures with intense pre-heating alters the evolution of the combustion process with respect to traditional flames, leading to features such as the susceptibility to oscillations, which are undesirable during combustion. Therefore, an effective use of advanced combustion technologies requires a thorough analysis of the combustion kinetic characteristics in order to identify optimal operating conditions and control strategies with high efficiency and low pollutant emissions. The present work experimentally and numerically characterized the ignition and oxidation processes of methane and propane, highly diluted in nitrogen, at atmospheric pressure, in a Plug Flow Reactor and a Perfectly Stirred Reactor under a wide range of operating conditions involving temperatures, mixture compositions and dilution levels. The attention was focused particularly on the chemistry of oscillatory phenomena and multistage ignitions. The global behavior of these systems can be qualitatively and partially quantitatively modeled using the detailed kinetic models available in the literature. Results suggested that, for diluted conditions and lower adiabatic flame temperatures, the competition among several pathways, i.e. intermediate- and

  9. Basic experiments during loss of vacuum event (LOVE) in fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ogawa, Masuro; Kunugi, Tomoaki; Seki, Yasushi (JAERI, Ibaraki (Japan))

    1993-06-01

    If a loss of vacuum event (LOVE) occurs due to damage of the vacuum vessel of a nuclear fusion experimental reactor, some chemical reactions such as a graphic oxidation and a buoyancy-driven exchange flow take place after equalization of the gas pressure between the inside and outside of the vacuum vessel. The graphite oxidation would generate inflammable carbon monoxide and release tritium retained in the graphite. The exchange flow through the breaches may transport the carbon monoxide and tritium out of the vacuum vessel. To add confidence to the safety evaluations and analyses, it is important to grasp the basic phenomena such as the exchange flow and the graphite oxidation. Experiments of the exchange flow and the graphite oxidation were carried out to obtain the exchange flow rate and the rate constant for the carbon monoxide combustion, respectively. These experimental results were compared with existing correlations. The authors plan a scaled-model test and a full-scale model test for the LOVE.

  10. Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2017-09-15

    In the paper measurement results of experimental modelling of a molten salt fast reactor concept will be presented and compared with three-dimensional computational fluid dynamics (CFD) simulation results. Purpose of this article is twofold, on one hand to introduce a geometry modification in order to avoid the disadvantages of the original geometry and discuss new measurement results. On the other hand to present an analysis in order to suggest a method of proper numerical modelling of the problem based on the comparison of calculation results and measurement data for the new, modified geometry. The investigated concept has a homogeneous cylindrical core without any internal structures. Previous measurements on the scaled and segmented plexiglas model of the concept core and simulation results have shown that this core geometry could be optimized for better thermal-hydraulics characteristics. In case of the original geometry strong undesired flow separation could develop, that could negatively affect the characteristics of the core from neutronics point of view as well. An internal flow distributor plate was designed and installed with the purpose of optimizing the flow field in the core by enhancing its uniformity. Particle image velocimetry (PIV) measurement results of the modified experimental model will be presented and compared to numerical simulation results with the purpose of CFD model validation.

  11. Experimental and numerical studies of microwave-plasma interaction in a MWPECVD reactor

    Directory of Open Access Journals (Sweden)

    A. Massaro

    2016-12-01

    Full Text Available This work deals with and proposes a simple and compact diagnostic method able to characterize the interaction between microwave and plasma without the necessity of using an external diagnostic tool. The interaction between 2.45 GHz microwave and plasma, in a typical ASTeX-type reactor, is investigated from experimental and numerical view points. The experiments are performed by considering plasmas of three different gas mixtures: H2, CH4-H2 and CH4-H2-N2. The two latter are used to deposit synthetic undoped and n-doped diamond films. The experimental setup equipped with a matching network enables the measurements of very low reflected power. The reflected powers show ripples due to the mismatching between wave and plasma impedance. Specifically, the three types of plasma exhibit reflected power values related to the variation of electron-neutral collision frequency among the species by changing the gas mixture. The different gas mixtures studied are also useful to test the sensitivity of the reflected power measurements to the change of plasma composition. By means of a numerical model, only the interaction of microwave and H2 plasma is examined allowing the estimation of plasma and matching network impedances and of reflected power that is found about eighteen times higher than that measured.

  12. Experimental investigations at the GENEVA passive residual heat removal test facility

    Energy Technology Data Exchange (ETDEWEB)

    Cloppenborg, Tim; Schuster, Christoph; Hurtado, Antonio [Technische Univ. Dresden (Germany). Professur fuer Wasserstoff- und Kernenergietechnik

    2014-07-01

    Phenomena of heat transfer system at low driving forces - mainly the transition zone between single phase and two phase heat transfer - is of high interest for several technical applications. Passive safety systems of advanced nuclear reactor concepts and operation of concentrated solar power systems are only two examples. The GENEVA natural circulation test facility was established for generic investigations of thermohydraulic impact factors on natural circulation residual heat removal systems at the Professorship of Hydrogen- and Nuclear Energy Technology, TU Dresden in 2013. (orig.)

  13. Study of process of water disinfection it saw energy solar using an experimental reactor; Estudo do proceso de desinfeccao de agua via energia solar utilizando um reator experimental

    Energy Technology Data Exchange (ETDEWEB)

    Batista, C. H.; Prado, L. R.; Lima, A. S.; Egues, S. M. S.; Araujo, P. M. M.

    2008-07-01

    In this work, was conducted an experimental study of the efficiency of a solar reactor in the disinfection of drinking water using photolysis (UV) and heterogeneous photo catalysis (TiO{sub 2}/UV). The experiments were conducted in batch mode, evaluating the effects of reactor inclination and the presence of a solar concentrator. The results indicated that the employed system was capable to promote the complete disinfection in 150 min using only the photo thermic effect, and in 120 min with the addition of immobilized TiO{sub 2} and the solar concentrator. (Author)

  14. Texas Experimental Tokamak, a plasma research facility: Technical progress report

    Energy Technology Data Exchange (ETDEWEB)

    Wootton, A.J.

    1995-08-01

    In the year just past, the authors made major progress in understanding turbulence and transport in both core and edge. Development of the capability for turbulence measurements throughout the poloidal cross section and intelligent consideration of the observed asymmetries, played a critical role in this work. In their confinement studies, a limited plasma with strong, H-mode-like characteristics serendipitously appeared and received extensive study though a diverted H-mode remains elusive. In the plasma edge, they appear to be close to isolating a turbulence drive mechanism. These are major advances of benefit to the community at large, and they followed from incremental improvements in diagnostics, in the interpretation of the diagnostics, and in TEXT itself. Their general philosophy is that the understanding of plasma physics must be part of any intelligent fusion program, and that basic experimental research is the most important part of any such program. The work here demonstrates a continuing dedication to the problems of plasma transport which continue to plague the community and are an impediment to the design of future devices. They expect to show here that they approach this problem consistently, systematically, and effectively.

  15. Efficient removal of antibiotics in a fluidized bed reactor by facile fabricated magnetic powdered activated carbon.

    Science.gov (United States)

    Ma, Jianqing; Yang, Qunfeng; Xu, Dongmei; Zeng, Xiaomei; Wen, Yuezhong; Liu, Weiping

    2017-02-01

    Powdered activated carbons (PACs) with micrometer size are showing great potential for enabling and improving technologies in water treatment. The critical problem in achieving practical application of PAC involves simple, effective fabrication of magnetic PAC and the design of a feasible reactor that can remove pollutants and recover the adsorbent efficiently. Herein, we show that such materials can be fabricated by the combination of PAC and magnetic Fe3O4 with chitosan-Fe hydrogel through a simple co-precipitation method. According to the characterization results, CS-Fe/Fe3O4/PAC with different micrometers in size exhibited excellent magnetic properties. The adsorption of tetracycline was fast and efficient, and 99.9% removal was achieved in 30 min. It also possesses good usability and stability to co-existing ions, organics, and different pH values due to its dispersive interaction nature. Finally, the prepared CS-Fe/Fe3O4/PAC also performed well in the fluidized bed reactor with electromagnetic separation function. It could be easily separated by applying a magnetic field and was effectively in situ regenerated, indicating a potential of practical application for the removal of pollutants from water.

  16. Estudio del comportamiento de reactores discontinuos y semicontinuos: modelización y comprobación experimental

    OpenAIRE

    Grau Vilalta, Ma. Dolors

    1999-01-01

    L'objectiu primordial d'aquest treball és la comparació entre el funcionament d'un reactor discontinu i un de semicontinu. Per això es porta a terme la modelització matemàtica d'ambdós, utilitzant programes propis emprant el llenguatge Fortran 77, a més del simulador ISIM i el software MATLAB. La validació dels models matemàtics s'efectua, en primer lloc, a partir de dades de la bibliografia. A partir d'aquí, es realitzen proves experimentals en una planta piloto amb un reactor encamisat de v...

  17. Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard Schultz

    2012-04-01

    Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

  18. Structural biology facilities at Brookhaven National Laboratory`s high flux beam reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korszun, Z.R.; Saxena, A.M.; Schneider, D.K. [Brookhaven National Laboratory, Upton, NY (United States)

    1994-12-31

    The techniques for determining the structure of biological molecules and larger biological assemblies depend on the extent of order in the particular system. At the High Flux Beam Reactor at the Brookhaven National Laboratory, the Biology Department operates three beam lines dedicated to biological structure studies. These beam lines span the resolution range from approximately 700{Angstrom} to approximately 1.5{Angstrom} and are designed to perform structural studies on a wide range of biological systems. Beam line H3A is dedicated to single crystal diffraction studies of macromolecules, while beam line H3B is designed to study diffraction from partially ordered systems such as biological membranes. Beam line H9B is located on the cold source and is designed for small angle scattering experiments on oligomeric biological systems.

  19. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  20. Experimental and statistical investigation of thermally induced failure in reactor fuel particles

    Energy Technology Data Exchange (ETDEWEB)

    Lunsford, J.L.; Imprescia, R.J.; Bowman, A.L.; Radosevich, C.E.

    1980-10-01

    An incomplete experimental study into the failure statistics of fuel particle for the high-temperature gas-cooled reactor (HTGR) is described. Fuel particles failure was induced by thermal ramping from room temperature to temperatures in the vicinity of 2273/sup 0/K to 2773/sup 0/K in 2 to 30 h and detected by the appearance of /sup 85/Kr in the helium carrier gas used to sweep the furnace. The concentration of krypton, a beta emitter, was detected by measuring the current that resulted when the helium sweep gas was passed through an ionization chamber. TRISO fuel particles gave a krypton concentration profile as a function of time that built up in several minutes and decayed in a fraction of an hour. This profile, which was temperature independent, was similar to the impulse response of the ionization chamber, suggesting that the TRISO particles failed instantaneously and completely. BISO fuel particles gave a krypton concentration profile as a function of time that built up in a fraction of an hour and decayed in a fraction of a day. This profile was strongly temperature dependent, suggesting that krypton release was diffusion controlled, i.e., that the krypton was diffusing through a sound coat, or that the BISO coating failed but that the krypton was unable to escape the kernel without diffusion, or that a combination of pre- and postfailure diffusion accompanied partial or complete failure.

  1. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Gauld, Ian C [ORNL

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  2. Fission product iodine release and retention in nuclear reactor accidents— experimental programme at PSI

    Science.gov (United States)

    Bruchertseifer, H.; Cripps, R.; Guentay, S.; Jaeckel, B.

    2003-01-01

    Iodine radionuclides constitute one of the most important fission products of uranium and plutonium. If the volatile forms would be released into the environment during a severe accident, a potential health hazard would then ensue. Understanding its behaviour is an important prerequisite for planning appropriate mitigation measures. Improved and extensive knowledge of the main iodine species and their reactions important for the release and retention processes in the reactor containment is thus mandatory. The aim of PSI's radiolytical studies is to improve the current thermodynamic and kinetic databases and the models for iodine used in severe accident computer codes. Formation of sparingly soluble silver iodide (AgI) in a PWR containment sump can substantially reduce volatile iodine fraction in the containment atmosphere. However, the effectiveness is dependent on its radiation stability. The direct radiolytic decomposition of AgI and the effect of impurities on iodine volatilisation were experimentally determined at PSI using a remote-controlled and automated high activity 188W/Re generator (40 GBq/ml). Low molecular weight organic iodides are difficult to be retained in engineered safety systems. Investigation of radiolytic decomposition of methyl iodide in aqueous solutions, combined with an on-line analysis of iodine species is currently under investigation at PSI.

  3. Design study of toroidal magnets for tokamak experimental power reactors. [NbTi alloys

    Energy Technology Data Exchange (ETDEWEB)

    Stekly, Z.J.J.; Lucas, E.J. (eds.)

    1976-12-01

    This report contains the results of a six-month study of superconducting toroidal field coils for a Tokamak Experimental Power Reactor to be built in the late 1980s. The designs are for 8 T and 12 T maximum magnetic field at the superconducting winding. At each field level two main concepts were generated; one in which each of the 16 coils comprising the system has an individual vacuum vessel and the other in which all the coils are contained in a single vacuum vessel. The coils have a D shape and have openings of 11.25 m x 7.5 m for the 8 T coils and 10.2 m x 6.8 m for the 12 T coils. All the designs utilize rectangular cabled conductor made from copper stabilized Niobium Titanium composite which operates at 4.2 K for the 8 T design and at 2.5 K for the 12 T design. Manufacturing procedures, processes and schedule estimates are also discussed.

  4. Bulk-bronzied graphites for plasma-facing components in ITER (International Thermonuclear Experimental Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Hirooka, Y.; Conn, R.W.; Doerner, R.; Khandagle, M. (California Univ., Los Angeles, CA (USA). Inst. of Plasma and Fusion Research); Causey, R.; Wilson, K. (Sandia National Labs., Livermore, CA (USA)); Croessmann, D.; Whitley, J. (Sandia National Labs., Albuquerque, NM (USA)); Holland, D.; Smolik, G. (Idaho National Engineering Lab., Idaho Falls, ID (USA)); Matsuda, T.; Sogabe, T. (Toyo Tanso Co. Ltd., O

    1990-06-01

    Newly developed bulk-boronized graphites and boronized C-C composites with a total boron concentration ranging from 1 wt % to 30 wt % have been evaluated as plasma-facing component materials for the International Thermonuclear Experimental Reactor (ITER). Bulk-boronized graphites have been bombarded with high-flux deuterium plasmas at temperatures between 200 and 1600{degree}C. Plasma interaction induced erosion of bulk-boronized graphites is observed to be a factor of 2--3 smaller than that of pyrolytic graphite, in regimes of physical sputtering, chemical sputtering and radiation enhanced sublimation. Postbombardment thermal desorption spectroscopy indicates that bulk-boronized graphites enhance recombinative desorption of deuterium, which leads to a suppression of the formation of deuterocarbon due to chemical sputtering. The tritium inventory in graphite has been found to decrease by an order of magnitude due to 10 wt % bulk-boronization at temperatures above 1000{degree}C. The critical heat flux to induce cracking for bulk-boronized graphites has been found to be essentially the same as that for non-boronized graphites. Also, 10 wt % bulk-boronization of graphite hinders air oxidation nearly completely at 800{degree}C and reduces the steam oxidation rate by a factor of 2--3 at around 1100 and 1350{degree}C. 38 refs., 5 figs.

  5. Economic impacts on the United States of siting decisions for the international thermonuclear experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Peerenboom, J.P.; Hanson, M.E.; Huddleston, J.R. [and others

    1996-08-01

    This report presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that, along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively.

  6. An experimental study on coolability through the external reactor vessel cooling according to RPV insulation design

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung Ho; Koo, Kil Mo; Park, Rae Joon; Cho, Young Ro; Kim, Sang Baik

    2004-01-01

    LAVA-ERVC experiments have been performed to investigate the effect of insulation design features on the water accessibility and coolability in case of the external reactor vessel cooling. Alumina iron thermite melt was used as corium stimulant. And the hemispherical test vessel is linearly scaled-down of RPV lower plenum. 4 tests have been performed varying the melt composition and the configuration of the insulation system. Due to the limited steam venting capacity through the insulation, steam binding occurred inside the annulus in the LAVA- ERVC-1, 2 tests which were performed for simulating the KSNP insulation design. This steam binding brought about incident heat up of the vessel outer surface at the upper part in the LAVA-ERVC-1, 2 tests. On the contrary, in the LAVA-ERVC-3, 4 tests which were performed for simulating the APR1400 insulation design, the temperatures of the vessel outer surface maintained near saturation temperature. Sufficient water ingression and steam venting through the insulation lead to effective cooldown of the vessel characterized by nucleate boiling in the LAVA-ERVC-3, 4 tests. From the LAVA-ERVC experimental results, it could be preliminarily concluded that if pertinent modification of the insulation design focused on the improvement of water ingression and steam venting should be preceded the possibility of in-vessel corium retention through the external vessel cooling could be considerably increased.

  7. A statistical experimental design to remove sulfate by crystallization in a fluidized-bed reactor

    Directory of Open Access Journals (Sweden)

    Mark Daniel G. de Luna

    2017-05-01

    Full Text Available This study used crystallization in a fluidized-bed reactor as an alternative technology to the conventional chemical precipitation to remove sulfate. The Box-Behnken Design was used to study the effects and interactions of seed dosage of synthetic gypsum, initial sulfate concentration and molar ratio of calcium to sulfate on conversion and removal of sulfate. The optimum conditions of conversion and removal of sulfate were determined and used to treat the simulated acid mine drainage (AMD wastewater. The effect of inorganic ions CO32−, NH4+ and Al3+ on sulfate conversion was also investigated. Experimental results indicated that seed dosage, initial sulfate concentration and molar ratio of calcium to sulfate are all significant parameters in the sulfate removal by fluidized-bed crystallization. The optimal conditions of 4 g seed L−1, 119.7 mM of initial sulfate concentration and [Ca2+]/[SO42−] molar ratio of 1.48 resulted in sulfate conversion of 82% and sulfate removal of 67%. Conversion and removal of sulfate in the simulated AMD wastewater were 79 and 63%, respectively. When ammonium or aluminum was added to the synthetic sulfate wastewater, significant conversion of sulfate was achieved.

  8. Economic Impacts on the United States of Siting Decisions for the International Thermonuclear Experimental Reactor

    Science.gov (United States)

    Peerenboom, J. P.; Hanson, M. E.; Huddleston, J. R.; Wolsko, T. D.

    1997-12-01

    This paper presents the results of a study that examines and compares the probable short-term economic impacts of the International Thermonuclear Experimental Reactor (ITER) on the United States (U.S.) if (1) ITER were to be sited in the U.S., or (2) ITER were to be sited in one of the other countries that, along with the U.S., is currently participating in the ITER program. Life-cycle costs associated with ITER construction, operation, and decommissioning are analyzed to assess their economic impact. A number of possible U.S. host and U.S. non-host technology and cost-sharing arrangements with the other ITER Parties are examined, although cost-sharing arrangements and the process by which the Parties will select a host country and an ITER site remain open issues. Both national and local/regional economic impacts, as measured by gross domestic product, regional output, employment, net exports, and income, are considered. These impacts represent a portion of the complex, interrelated set of economic considerations that characterize U.S. host and U.S. non-host participation in ITER. A number of other potentially important economic and noneconomic considerations are discussed qualitatively.

  9. Conceptual studies of toroidal field magnets for the tokamak (fusion) experimental power reactor. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1976-11-01

    This report presents the results of ''Conceptual Studies of Toroidal Field Magnets for the Tokamak Experimental Power Reactor'' performed for the Energy Research and Development Administration, Oak Ridge Operations. Two conceptual coil designs are developed. One design approach to produce a specified 8 Tesla maximum field uses a novel NbTi superconductor design cooled by pool-boiling liquid helium. For a highest practicable field design, a unique NbSn/sub 3/ conductor is used with forced-flow, single-phase liquid helium cooling to achieve a 12 Tesla peak field. Fabrication requirements are also developed for these approximately 7 meter horizontal bore by 11 meter vertical bore coils. Cryostat design approaches are analyzed and a hybrid cryostat approach selected. Structural analyses are performed for approaches to support in-plane and out-of-plane loads and a structural approach selected. In addition to the conceptual design studies, cost estimates and schedules are prepared for each of the design approaches, major uncertainties and recommendations for research and development identified, and test coil size for demonstration recommended.

  10. Risk-Informing Safety Reviews for Non-Reactor Nuclear Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Mubayi, V.; Azarm, A.; Yue, M.; Mukaddam, W.; Good, G.; Gonzalez, F.; Bari, R.A.

    2011-03-13

    This paper describes a methodology used to model potential accidents in fuel cycle facilities that employ chemical processes to separate and purify nuclear materials. The methodology is illustrated with an example that uses event and fault trees to estimate the frequency of a specific energetic reaction that can occur in nuclear material processing facilities. The methodology used probabilistic risk assessment (PRA)-related tools as well as information about the chemical reaction characteristics, information on plant design and operational features, and generic data about component failure rates and human error rates. The accident frequency estimates for the specific reaction help to risk-inform the safety review process and assess compliance with regulatory requirements.

  11. Transmission thermography for inspecting the busbar insulation layer in thermonuclear experimental reactor

    Science.gov (United States)

    Chen, Dapeng; Zhang, Guang; Zhang, Xiaolong; Zeng, Zhi

    2014-11-01

    In Thermonuclear Experimental Reactor, Superconducting Busbar is used for current transmission between magnet coils and current leads. The work temperature of the Busbar is about 4K because of liquid helium via inside. The large temperature grad from 300K to 4K could lead to the defects and damages occur on the insulation layer, which is made of glass fiber and polyimide and has a big different thermal expansion coefficient compared with the metal inner cylinder. This paper aims at developing an infrared transmission non-destructive evaluation (NDE) method for inspecting the insulation layer of Superconducting Busbar; theoretical model of transient heat conduction under a continuous inner heat source for cylindrical structure is described in the paper; a Busbar specimen which is designed with three delamination defects of different depths is heated inside by pouring hot water and monitored by an infrared detector located outside. Results demonstrate excellent detection performance for delamination defects in the insulation layer by using transmission thermography, all of the three defects of different depths can be visualized clearly in the thermal images, and the deeper defect has a better signal contrast, which is also shown in the temperature difference between defects and sound area vs. time curves. The results of light pulse thermography is also shown as a comparison, and it is found that the thermal images obtained by the transmission thermography has a much better signal contrast than that of the pulse thermography. In order to verify the experiments, finite element method is applied to simulate the heat conduction in the Busbar under the continuous inside heating, and it is found that the simulated temperature vs. time and simulated temperature difference vs. time curves are basically coincident with the experimental results. In addition, the possibility of in-service inspection for Busbar insulation layer in ITER item is discussed.

  12. A novel reverse flow reactor coupling endothermic and exothermic reactions: an experimental study

    NARCIS (Netherlands)

    van Sint Annaland, M.; Nijssen, R.C.

    2002-01-01

    A new reactor concept is studied for highly endothermic heterogeneously catalysed gas phase reactions at high temperatures with rapid but reversible catalyst deactivation. The reactor concept aims to achieve an indirect coupling of energy necessary for endothermic reactions and energy released by

  13. Experimental and numerical stability investigations on natural circulation boiling water reactors

    CERN Document Server

    Marcel, CP

    2007-01-01

    In the design of novel nuclear reactors active systems are replaced by passive ones in order to reduce the risk of failure. For that reason natural circulation is being considered as the primary cooling mechanism in next generation nuclear reactor designs

  14. Participation in the US Department of Energy reactor sharing program. Final progress report, October 1996--September 1997

    Energy Technology Data Exchange (ETDEWEB)

    Mulder, R.U.; Benneche, P.E.; Hosticka, B.

    1998-04-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA from use at their institutions. These areas are discussed in this report.

  15. Participation in the US Department of Energy Reactor Sharing Program. Annual report, September 30, 1993--September 29, 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    The objective of the DOE supported Reactor Sharing Program is to increase the availability of university nuclear reactor facilities to non-reactor-owning educational institutions. The educational and research programs of these user institutions is enhanced by the use of the nuclear facilities. Several methods have been used by the UVA Reactor Facility to achieve this objective. First, many college and secondary school groups toured the Reactor Facility and viewed the UVAR reactor and associated experimental facilities. Second, advanced undergraduate and graduate classes from area colleges and universities visited the facility to perform experiments in nuclear engineering and physics which would not be possible at the user institution. Third, irradiation and analysis services at the Facility have been made available for research by faculty and students from user institutions. Fourth, some institutions have received activated material from UVA for use at their institutions. These areas are discussed further in the report.

  16. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus - Spectral indices

    Science.gov (United States)

    Perret, G.; Pattupara, R. M.; Girardin, G.; Chawla, R.

    2013-03-01

    PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970's to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices - including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%).

  17. NASA's Nuclear Frontier: The Plum Brook Reactor Facility, 1941-2002

    Science.gov (United States)

    Bowles, Mark D.; Arrighi, Robert S.

    2004-01-01

    In 1953, President Eisenhower delivered a speech called "Atoms for Peace" to the United Nations General Assembly. He described the emergence of the atomic age and the weapons of mass destruction that were piling up in the storehouses of the American and Soviet nations. Although neither side was aiming for global destruction, Eisenhower wanted to "move out of the dark chambers of horrors into the light, to find a way by which the minds of men, the hopes of men, the souls of men everywhere, can move towards peace and happiness and well-being." One way Eisenhower hoped this could happen was by transforming the atom from a weapon of war into a useful tool for civilization. Many people believed that there were unprecedented opportunities for peaceful nuclear applications. These included hopeful visions of atomic-powered cities, cars, airplanes, and rockets. Nuclear power might also serve as an efficient way to generate electricity in space to support life and machines. Eisenhower wanted to provide scientists and engineers with "adequate amounts of fission- able material with which to test and develop their ideas." But, in attempting to devise ways to use atomic power for peaceful purposes, scientists realized how little they knew about the nature and effects of radiation. As a result, the United States began constructing nuclear test reactors to enable scientists to conduct research by producing neutrons.

  18. Report on emergency electrical power supply systems for nuclear fuel cycle and reactor facilities security systems

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    The report includes information that will be useful to those responsible for the planning, design and implementation of emergency electric power systems for physical security and special nuclear materials accountability systems. Basic considerations for establishing the system requirements for emergency electric power for security and accountability operations are presented. Methods of supplying emergency power that are available at present and methods predicted to be available in the future are discussed. The characteristics of capacity, cost, safety, reliability and environmental and physical facility considerations of emergency electric power techniques are presented. The report includes basic considerations for the development of a system concept and the preparation of a detailed system design.

  19. Experimental and Theoretical Progress of Linear Collider Final Focus Design and ATF2 Facility

    CERN Document Server

    Seryi, Andrei; Zimmermann, Frank; Kubo, Kiyoshi; Kuroda, Shigeru; Okugi, Toshiyuki; Tauchi, Toshiaki; Terunuma, Nobuhiro; Urakawa, Junji; White, Glen; Woodley, Mark; Angal-Kalinin, Deepa

    2014-01-01

    In this brief overview we will reflect on the process of the design of the linear collider (LC) final focus (FF) optics, and will also describe the theoretical and experimental efforts on design and practical realisation of a prototype of the LC FF optics implemented in the ATF2 facility at KEK, Japan, presently being commissioned and operated.

  20. Experimental data from a full-scale facility investigating radiant and convective terminals

    DEFF Research Database (Denmark)

    Le Dreau, Jerome; Heiselberg, Per; Jensen, Rasmus Lund

    The objective of this technical report is to provide information on the accuracy of the experiments performed in “the Cube” (part I, II and III). Moreover, this report lists the experimental data, which have been monitored in the test facility (part IV). These data are available online and can be...

  1. Ecological Realism of U.S. EPA Experimental Stream Facility Studies

    Science.gov (United States)

    The USEPA’s Experimental Stream Facility (ESF) conducts meso-scale ecotoxicology studies that account for both structural and functional responses of whole stream communities to contaminants or other stressors. The 16 mesocosms of ESF are indoors and consist of a tiled run sectio...

  2. Upgrade of U.S. EPA's Experimental Stream Facility Supervisory Control and Data Acquisition System

    Science.gov (United States)

    The Supervisory control and data acquisition (SCADA) system for the U.S. EPA’s Experimental Stream Facility (ESF) was upgraded using Camile hardware and software in 2015. The upgrade added additional hardwired connections, new wireless capabilities, and included a complete rewrit...

  3. Hydrocarbon pyrolysis reactor experimentation and modeling for the production of solar absorbing carbon nanoparticles

    Science.gov (United States)

    Frederickson, Lee Thomas

    Much of combustion research focuses on reducing soot particulates in emissions. However, current research at San Diego State University (SDSU) Combustion and Solar Energy Laboratory (CSEL) is underway to develop a high temperature solar receiver which will utilize carbon nanoparticles as a solar absorption medium. To produce carbon nanoparticles for the small particle heat exchange receiver (SPHER), a lab-scale carbon particle generator (CPG) has been built and tested. The CPG is a heated ceramic tube reactor with a set point wall temperature of 1100-1300°C operating at 5-6 bar pressure. Natural gas and nitrogen are fed to the CPG where natural gas undergoes pyrolysis resulting in carbon particles. The gas-particle mixture is met downstream with dilution air and sent to the lab scale solar receiver. To predict soot yield and general trends in CPG performance, a model has been setup in Reaction Design CHEMKIN-PRO software. One of the primary goals of this research is to accurately measure particle properties. Mean particle diameter, size distribution, and index of refraction are calculated using Scanning Electron Microscopy (SEM) and a Diesel Particulate Scatterometer (DPS). Filter samples taken during experimentation are analyzed to obtain a particle size distribution with SEM images processed in ImageJ software. These results are compared with the DPS, which calculates the particle size distribution and the index of refraction from light scattering using Mie theory. For testing with the lab scale receiver, a particle diameter range of 200-500 nm is desired. Test conditions are varied to understand effects of operating parameters on particle size and the ability to obtain the size range. Analysis of particle loading is the other important metric for this research. Particle loading is measured downstream of the CPG outlet and dilution air mixing point. The air-particle mixture flows through an extinction tube where opacity of the mixture is measured with a 532 nm

  4. Parametric experimental tests of steam gasification of pine wood in a fluidized bed reactor

    Directory of Open Access Journals (Sweden)

    L. Vecchione

    2013-09-01

    Full Text Available Among Renewable Energy Sources (RES, biomass represent one of the most common and suitable solution in order to contribute to the global energy supply and to reduce greenhouse gases (GHG emissions. The disposal of some residual biomass, as pruning from pine trees, represent a problem for agricultural and agro-industrial sectors. But if the residual biomass are used for energy production can become a resource. The most suitable energy conversion technology for the above-mentioned biomass is gasification process because the high C/N ratio and the low moisture content, obtained from the analysis. In this work a small-pilot bubbling-bed gasification plant has been designed, constructed and used in order to obtain, from the pine trees pruning, a syngas with low tar and char contents and high hydrogen content. The activities showed here are part of the activities carried out in the European 7FP UNIfHY project. In particular the aim of this work is to develop experimental test on a bench scale steam blown fluidized bed biomass gasifier. These tests will be utilized in future works for the simulations of a pilot scale steam fluidized bed gasifier (100 kWth fed with different biomass feedstock. The results of the tests include produced gas and tar composition as well gas, tar and char yield. Tests on a bench scale reactor (8 cm I.D. were carried out varying steam to biomass ratio from 0.5, 0.7 and 1 to 830°C.

  5. Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 2008

    Energy Technology Data Exchange (ETDEWEB)

    U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research

    2009-12-01

    This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 2008 annual reports submitted by five of the seven categories1 of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. The annual reports submitted by these licensees consist of radiation exposure records for each monitored individual. These records are analyzed for trends and presented in this report in terms of collective dose and the distribution of dose among the monitored individuals. Because there are no geologic repositories for high-level waste currently licensed and no low-level waste disposal facilities in operation, only five categories will be considered in this report.

  6. GEANT4 used for neutron beam design of a neutron imaging facility at TRIGA reactor in Morocco

    Science.gov (United States)

    Ouardi, A.; Machmach, A.; Alami, R.; Bensitel, A.; Hommada, A.

    2011-09-01

    Neutron imaging has a broad scope of applications and has played a pivotal role in visualizing and quantifying hydrogenous masses in metallic matrices. The field continues to expand into new applications with the installation of new neutron imaging facilities. In this scope, a neutron imaging facility for computed tomography and real-time neutron radiography is currently being developed around 2.0MW TRIGA MARK-II reactor at Maamora Nuclear Research Center in Morocco (Reuscher et al., 1990 [1]; de Menezes et al., 2003 [2]; Deinert et al., 2005 [3]). The neutron imaging facility consists of neutron collimator, real-time neutron imaging system and imaging process systems. In order to reduce the gamma-ray content in the neutron beam, the tangential channel was selected. For power of 250 kW, the corresponding thermal neutron flux measured at the inlet of the tangential channel is around 3×10 11 ncm 2/s. This facility will be based on a conical neutron collimator with two circular diaphragms with diameters of 4 and 2 cm corresponding to L/D-ratio of 165 and 325, respectively. These diaphragms' sizes allow reaching a compromise between good flux and efficient L/D-ratio. Convergent-divergent collimator geometry has been adopted. The beam line consists of a gamma filter, fast neutrons filter, neutron moderator, neutron and gamma shutters, biological shielding around the collimator and several stages of neutron collimator. Monte Carlo calculations by a fully 3D numerical code GEANT4 were used to design the neutron beam line ( http://www.info.cern.ch/asd/geant4/geant4.html[4]). To enhance the neutron thermal beam in terms of quality, several materials, mainly bismuth (Bi) and sapphire (Al 2O 3) were examined as gamma and neutron filters respectively. The GEANT4 simulations showed that the gamma and epithermal and fast neutron could be filtered using the bismuth (Bi) and sapphire (Al 2O 3) filters, respectively. To get a good cadmium ratio, GEANT 4 simulations were used to

  7. Fast reactor: an experimental study of thermohydraulic processes in different operating regimes

    Science.gov (United States)

    Opanasenko, A. N.; Sorokin, A. P.; Zaryugin, D. G.; Trufanov, A. A.

    2017-05-01

    Results of integrated water model studies of temperature fields and a flow pattern of a nonisothermal primary coolant in the elements of the fast neutron reactor (hereinafter, fast reactor) primary circuit with primary sodium in different regimes, such as forced circulation (FC), transition to the reactor cooldown and emergency cooldown with natural coolant convection, are presented. It is shown that, under the influence of lift forces on the nonisothermal coolant flow in the upper chamber at the periphery of its bottom region over the side shields, a stable cold coolant isothermal zone is formed, whose dimensions increase with increase of total water flowrate. An essential and stable coolant temperature stratification is detected in the peripheral area of the upper (hot) chamber over the side shields, in the pressure and cold side chambers, in the elevator baffle, in the cooling system of the reactor vessel, and in the outlet of intermediate and autonomous heat exchangers in different operating regimes. Large gradients and temperature fluctuations are registered at the interface of stratified and recycling formations. In all of the studied cooldown versions, the coolant outlet temperature at the core fuel assembly is decreased and the coolant temperature in the peripheral zone of the upper chamber is increased compared to the FC. High performance of a passive emergency cooldown system of a fast reactor (BN-1200) with submersible autonomous heat exchangers (AHE) is confirmed. Thus, in a normal operation regime, even in case of malfunction of three submersible AHEs, the temperature of the equipment inside the reactor remains within acceptable limits and decay heat removal from the reactor does not exceed safe operation limits. The obtained results can be used both for computer code verification and for approximate estimate of the reactor plant parameters on the similarity criteria basis.

  8. Experimental evidences of95 mTc production in a nuclear reactor.

    Science.gov (United States)

    Cohen, I M; Robles, A; Mendoza, P; Airas, R M; Montoya, E H

    2018-02-02

    95 m Tc has been identified as by-product in some solutions of 99 m Tc obtained by irradiation of molybdenum trioxide in a reactor neutron flux. The characterization was carried out using both measurements by gamma spectrometry and half-life determination. The possible ways that lead to the 95 m Tc production in a nuclear reactor are discussed. Copyright © 2018. Published by Elsevier Ltd.

  9. Experimental evaluation of gamma fluence-rate predictions from Argon-41 releases to the atmosphere over a nuclear research reactor site

    DEFF Research Database (Denmark)

    Rojas-Palma, C.; Aage, H.K.; Astrup, P.

    2004-01-01

    An experimental study of radionuclide dispersion in the atmosphere has been conducted at the BR1 research reactor in Mol, Belgium. Artificially generated aerosols ('white smoke') were mixed with the routine releases of Ar-41 in the reactor's 60-m tall venting stack. The detailed plume geometry...

  10. 76 FR 11291 - University of New Mexico AGN-201M Reactor Notice of Issuance of Renewed Facility Operating...

    Science.gov (United States)

    2011-03-01

    ..., New Mexico. The UNMR is a solid homogeneous core research reactor licensed to operate at a steady... Nuclear Regulatory Commission. Jessie Quichocho, Chief, Research and Test Reactors Licensing Branch, Division of Policy and Rulemaking, Office of Nuclear Reactor Regulation. BILLING CODE 7590-01-P ...

  11. Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Kurt R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

  12. Los Alamos experimental capabilities: Ancho Canyon high explosives and pulse power facilities

    Energy Technology Data Exchange (ETDEWEB)

    Morris, C.E.

    1993-01-01

    This document outlines the Ancho Canyon testing facility comprehensive material characterization capabilities. These include the high explosive (HE) firing sites, a full complement of gun facilities, and variety of pulse power capacitor bank systems of various energies. The explosive fabrication capability at Los Alamos allows the design and testing of unique HE experimental assemblies. Depending on the hydrodynamic requirements, these explosive systems can vary widely in cost. Years of experience have enabled the development of a comprehensive set of diagnostics to monitor these experiments.

  13. Los Alamos experimental capabilities: Ancho Canyon high explosives and pulse power facilities

    Energy Technology Data Exchange (ETDEWEB)

    Morris, C.E.

    1993-02-01

    This document outlines the Ancho Canyon testing facility comprehensive material characterization capabilities. These include the high explosive (HE) firing sites, a full complement of gun facilities, and variety of pulse power capacitor bank systems of various energies. The explosive fabrication capability at Los Alamos allows the design and testing of unique HE experimental assemblies. Depending on the hydrodynamic requirements, these explosive systems can vary widely in cost. Years of experience have enabled the development of a comprehensive set of diagnostics to monitor these experiments.

  14. MaRIE: an experimental facility concept revolutionizing materials in extremes

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, Cris W [Los Alamos National Laboratory

    2011-01-07

    The Matter-Radiation Interactions in Extremes (MaRIE) project intends to create an experimental facility that will revolutionize the control of materials in extremes. That control extends to extreme regimes where solid material has failed and begins to flow - the regimes of fluid dynamics and turbulent mixing. This presentation introduces the MaRIE facility concept, demonstrates examples of the science case that determine its functional requirements, and kicks-off the discussion of the decadal scientific challenges of mixing in extremes, including those MaRIE might address.

  15. SDG and E - ERDA Geothermal Loop Experimental Facility. Bi-monthly report, May 1975-August 1976

    Energy Technology Data Exchange (ETDEWEB)

    Bishop, H.K.; Bricarello, J.R.; Campbell, J.A.; Lombard, G.L.; Mulliner, D.K.; Swanson, C.R.

    1976-09-01

    A description of the Geothermal Loop Experimental Facility (GLEF) its construction problems, and a discussion of start-up testing are included. A history and description of the operation and maintenance with the brine injection pump for the facility are presented. The GLEF was divided into five separate sections: steam and condensate system, brine system, purge water system, vent gas system, and cooling water system. An insight into the chemistry of each system is provided by analysis of samples taken. Scaling and corrosion effects of brine, steam, gas, and water in these systems are described in detail. (MHR)

  16. Advanced Safeguards Approaches for New Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  17. Equipment for neutron measurements at VR-1 Sparrow training reactor.

    Science.gov (United States)

    Kolros, Antonin; Huml, Ondrej; Kríz, Martin; Kos, Josef

    2010-01-01

    The VR-1 sparrow reactor is an experimental nuclear facility for training, student education and teaching purposes. The sparrow reactor is an educational platform for the basic experiments at the reactor physic and dosimetry. The aim of this article is to describe the new experimental equipment EMK310 features and possibilities for neutron detection by different gas filled detectors at VR-1 reactor. Among the EMK310 equipment typical attributes belong precise set-up, simple control, resistance to electromagnetic interference, high throughput (counting rate), versatility and remote controllability. The methods for non-linearity correction of pulse neutron detection system and reactimeter application are presented. Copyright 2009. Published by Elsevier Ltd.

  18. Technical specifications, Hanford production reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gilbert, W.D. [comp.

    1962-06-25

    These technical specifications are applicable to the eight operating production reactor facilities, B, C, D, DR, F, H, KE, and KW. Covered are operating and performance restrictions and administrative procedures. Areas covered by the operating and performance restrictions are reactivity, reactor control and safety elements, power level, temperature and heat flux, reactor fuel loadings, reactor coolant systems, reactor confinement, test facilities, code compliance, and reactor scram set points. Administrative procedures include process control procedures, training programs, audits and inspections, and reports and records.

  19. Reanalysis of the Gas-cooled fast reactor experiments at the zero power facility Proteus – Spectral indices

    Directory of Open Access Journals (Sweden)

    Girardin G.

    2013-03-01

    Full Text Available PROTEUS is a zero power reactor at the Paul Scherrer Institute which has been employed during the 1970’s to study experimentally the physics of the gas-cooled fast reactor. Reaction rate distributions, flux spectrum and reactivity effects have been measured in several configurations featuring PuO2/UO2 fuel, absorbers, large iron shields, and thorium oxide and thorium metal fuel either distributed quasihomogeneously in the reference PuO2/UO2 lattice or introduced in the form of radial and axial blanket zones. This papers focus on the spectral indices – including fission and capture in 232Th and 237Np - measured in the reference PuO2/UO2 lattices and their predictions with an MCNPX model specially developed for the PROTEUS-GCFR core. Predictions were obtained with JEFF-3.1 and -3.11, ENDF/B-VII.0 and VII.1, and JENDL-3.3 and -4.0. A general good agreement was demonstrated. The ratio of 232Th fission to 239Pu fission, however, was under-predicted by 8.7±2.1% and 6.5±2.1% using ENDF/B-VII.0 and VII.1, respectively. Finally, the capture rates in 237Np tended to be underpredicted by the JEFF and JENDL libraries, although the new cross section in JEFF-3.1.1 slightly improved the 237Np capture to 239Pu fission results (3.4±2.4%.

  20. Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S.T. Revankar; W. Zhou; Gavin Henderson

    2008-07-08

    The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

  1. Research on accelerator-driven transmutation and studies of experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Takizuka, Takakazu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    JAERI is carrying out R and Ds on accelerator-driven transmutation systems under the national OMEGA Program that aims at development of the technology to improve efficiency and safety in the final disposal of radioactive waste. Research facilities for accelerator-driven transmutation experiments are proposed to construct within the framework of the planned JAERI Neutron Science Project. This paper describes the features of the proposed accelerator-driven transmutation systems and their technical issues to be solved. A research facility plan under examination is presented. The plan is divided in two phases. In the second phase, technical feasibility of accelerator-driven systems will be demonstrated with a 30-60 MW experimental integrated system and with a 7 MW high-power target facility. (author)

  2. Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kalimullah, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Olson, Arne P. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. E. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanan, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Dionne, B. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-04-07

    The document compiles in a single volume several verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. Some works that are available in the open literature are simply referenced at the outset, and are not included in the document. PLTEMP has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given. Each chapter of the document deals with the verification or validation of a specific model. The model verification is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The model validation is done by comparing the code with experimental data or a more validated code like the RELAP5 code.

  3. Screening of Gas-Cooled Reactor Thermal-Hydraulic and Safety Analysis Tools and Experimental Database

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Min Hwan; Lee, Seung Wook (and others)

    2007-08-15

    This report is a final report of I-NERI Project, 'Screening of Gas-cooled Reactor Thermal Hydraulic and Safety Analysis Tools and Experimental Database 'jointly carried out by KAERI, ANL and INL. In this study, we developed the basic technologies required to develop and validate the VHTR TH/safety analysis tools and evaluated the TH/safety database information. The research tasks consist of; 1) code qualification methodology (INL), 2) high-level PIRTs for major nucleus set of events (KAERI, ANL, INL), 3) initial scaling and scoping analysis (ANL, KAERI, INL), 4) filtering of TH/safety tools (KAERI, INL), 5) evaluation of TH/safety database information (KAERI, INL, ANL) and 6) key scoping analysis (KAERI). The code qualification methodology identifies the role of PIRTs in the R and D process and the bottom-up and top-down code validation methods. Since the design of VHTR is still evolving, we generated the high-level PIRTs referencing 600MWth block-type GT-MHR and 400MWth pebble-type PBMR. Nucleus set of events that represents the VHTR safety and operational transients consists of the enveloping scenarios of HPCC (high pressure conduction cooling: loss of primary flow), LPCC/Air-Ingress (low pressure conduction cooling: loss of coolant), LC (load changes: power maneuvering), ATWS (anticipated transients without scram: reactivity insertion), WS (water ingress: water-interfacing system break) and HU (hydrogen-side upset: loss of heat sink). The initial scaling analysis defines dimensionless parameters that need to be reflected in mixed convection modeling and the initial scoping analysis provided the reference system transients used in the PIRTs generation. For the PIRTs phenomena, we evaluated the modeling capability of the candidate TH/safety tools and derived a model improvement need. By surveying and evaluating the TH/safety database information, a tools V and V matrix has been developed. Through the key scoping analysis using available database, the

  4. Measurement of the 240Pu(n,f cross-section at the CERN n_TOF facility: First results from experimental area II (EAR-2

    Directory of Open Access Journals (Sweden)

    Stamatopoulos A.

    2017-01-01

    Full Text Available The accurate knowledge of the neutron-induced fission cross-sections of actinides and other isotopes involved in the nuclear fuel cycle is essential for the design of advanced nuclear systems, such as Generation-IV nuclear reactors. Such experimental data can also provide the necessary feedback for the adjustment of nuclear model parameters used in the evaluation process, resulting in the further development of nuclear fission models. In the present work, the 240Pu(n,f cross-section was measured at CERN's n_TOF facility relative to the well-known 235U(n,f cross section, over a wide range of neutron energies, from meV to almost MeV, using the time-of-flight technique and a set-up based on Micromegas detectors. This measurement was the first experiment to be performed at n_TOF's new experimental area (EAR-2, which offers a significantly higher neutron flux compared to the already existing experimental area (EAR-1. Preliminary results as well as the experimental procedure, including a description of the facility and the data handling and analysis, are presented.

  5. Measurement of the 240Pu(n,f) cross-section at the CERN n_TOF facility: First results from experimental area II (EAR-2)

    Science.gov (United States)

    Stamatopoulos, A.; Tsinganis, A.; Colonna, N.; Vlastou, R.; Kokkoris, M.; Schillebeeckx, P.; Plompen, A.; Heyse, J.; Žugec, P.; Barbagallo, M.; Calviani, M.; Berthoumieux, E.; Chiaveri, E.; Aberle, O.; Andrzejewski, J.; Audouin, L.; Bécares, V.; Bacak, M.; Balibrea, J.; Barros, S.; Bečvář, F.; Beinrucker, C.; Belloni, F.; Billowes, J.; Boccone, V.; Bosnar, D.; Brugger, M.; Caamaño, M.; Calviño, F.; Cano-Ott, D.; Cerutti, F.; Cortés, G.; Cortés-Giraldo, M. A.; Cosentino, L.; Damone, L. A.; Deo, K.; Diakaki, M.; Domingo-Pardo, C.; Dressler, R.; Dupont, E.; Durán, I.; Fernández-Domínguez, B.; Ferrari, A.; Ferreira, P.; Finocchiaro, P.; Frost, R. J. W.; Furman, V.; Göbel, K.; Gómez-Hornillos, M. B.; García, A. R.; Gheorghe, I.; Glodariu, T.; Gonçalves, I. F.; González, E.; Goverdovski, A.; Griesmayer, E.; Guerrero, C.; Gunsing, F.; Harada, H.; Heftrich, T.; Heinitz, S.; Hernández-Prieto, A.; Jenkins, D. G.; Jericha, E.; Käppeler, F.; Kadi, Y.; Katabuchi, T.; Kavrigin, P.; Ketlerov, V.; Khryachkov, V.; Kimura, A.; Kivel, N.; Krtička, M.; Leal-Cidoncha, E.; Lederer, C.; Leeb, H.; Lerendegui-Marco, J.; Licata, M.; Meo, S. Lo; Losito, R.; Macina, D.; Marganiec, J.; Martínez, T.; Massimi, C.; Mastinu, P.; Mastromarco, M.; Matteucci, F.; Mendoza, E.; Mengoni, A.; Milazzo, P. M.; Mingrone, F.; Mirea, M.; Montesano, S.; Musumarra, A.; Nolte, R.; Palomo-Pinto, F. R.; Paradela, C.; Patronis, N.; Pavlik, A.; Perkowski, J.; Porras, J. I.; Praena, J.; Quesada, J. M.; Rauscher, T.; Reifarth, R.; Riego-Perez, A.; Robles, M.; Rubbia, C.; Ryan, J. A.; Sabaté-Gilarte, M.; Saxena, A.; Schmidt, S.; Schumann, D.; Sedyshev, P.; Smith, A. G.; Suryanarayana, S. V.; Tagliente, G.; Tain, J. L.; Tarifeño-Saldivia, A.; Tassan-Got, L.; Valenta, S.; Vannini, G.; Variale, V.; Vaz, P.; Ventura, A.; Vlachoudis, V.; Wallner, A.; Warren, S.; Weigand, M.; Weiss, C.; Wright, T.

    2017-09-01

    The accurate knowledge of the neutron-induced fission cross-sections of actinides and other isotopes involved in the nuclear fuel cycle is essential for the design of advanced nuclear systems, such as Generation-IV nuclear reactors. Such experimental data can also provide the necessary feedback for the adjustment of nuclear model parameters used in the evaluation process, resulting in the further development of nuclear fission models. In the present work, the 240Pu(n,f) cross-section was measured at CERN's n_TOF facility relative to the well-known 235U(n,f) cross section, over a wide range of neutron energies, from meV to almost MeV, using the time-of-flight technique and a set-up based on Micromegas detectors. This measurement was the first experiment to be performed at n_TOF's new experimental area (EAR-2), which offers a significantly higher neutron flux compared to the already existing experimental area (EAR-1). Preliminary results as well as the experimental procedure, including a description of the facility and the data handling and analysis, are presented.

  6. Preliminary Feasibility, Design, and Hazard Analysis of a Boiling Water Test Loop Within the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility

    Energy Technology Data Exchange (ETDEWEB)

    Douglas M. Gerstner

    2009-05-01

    The Advanced Test Reactor (ATR) is a pressurized light-water reactor with a design thermal power of 250 MW. The principal function of the ATR is to provide a high neutron flux for testing reactor fuels and other materials. The ATR and its support facilities are located at the Idaho National Laboratory (INL). A Boiling Water Test Loop (BWTL) is being designed for one of the irradiation test positions within the. The objective of the new loop will be to simulate boiling water reactor (BWR) conditions to support clad corrosion and related reactor material testing. Further it will accommodate power ramping tests of candidate high burn-up fuels and fuel pins/rods for the commercial BWR utilities. The BWTL will be much like the pressurized water loops already in service in 5 of the 9 “flux traps” (region of enhanced neutron flux) in the ATR. The loop coolant will be isolated from the primary coolant system so that the loop’s temperature, pressure, flow rate, and water chemistry can be independently controlled. This paper presents the proposed general design of the in-core and auxiliary BWTL systems; the preliminary results of the neutronics and thermal hydraulics analyses; and the preliminary hazard analysis for safe normal and transient BWTL and ATR operation.

  7. Experimental investigation into fast pyrolysis of biomass using an entrained flow reactor

    Science.gov (United States)

    Bohn, M.; Benham, C.

    1981-02-01

    Pyrolysis experiments were performed with steam as a carrier gas and two different feedstocks - wheat straw and powdered material derived from municipal solid waste (ECO-II TM). Reactor wall temperature was varied from 7000 to 1400 C. Gas composition data from the ECO-II tests were comparable to previously reported data but ethylene yield appeared to vary with reactor wall temperature and residence time. The important conclusion from the wheat straw tests is that olefin yields are about one half that obtained from ECO-II. Evidence was found that high olefin yields from ECO-II are due to the presence of plastics in the feedstock.

  8. Evaluation of activation detectors for the SPHINX project at the LR-0 experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lahodova, Zdena; Viererbl, Ladislav [Research Center Rez Ltd (Czech Republic); Novak, Evzen; Svadlenkova, Marie; Rypar, Vojtech [Nuclear Power and Safety Division, Nuclear Research Institute Rez plc (Czech Republic)

    2008-07-01

    This article summarizes the measurements of neutron fluence distributions carried out at the LR-0 research reactor (Czech Republic) in the frame of the SPHINX project. The influence of fluoride-salts or graphite filling in the SR-0 modules on neutron spectrum was studied using activation detectors. The activation detectors (Mn, Ni, In and Au) were evaluated to determine the changes in neutron field. The In and Au detectors were also irradiated with a cadmium cover. Five different configurations of reactor core (EROS) were realized. (authors)

  9. Experimental evaluation of methane dry reforming process on a membrane reactor to hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Fabiano S.A.; Benachour, Mohand; Abreu, Cesar A.M. [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil). Dept. of Chemical Engineering], Email: f.aruda@yahoo.com.br

    2010-07-01

    In a fixed bed membrane reactor evaluations of methane-carbon dioxide reforming over a Ni/{gamma}- Al{sub 2}O{sub 3} catalyst were performed at 773 K, 823 K and 873 K. A to convert natural gas into syngas a fixed-bed reactor associate with a selective membrane was employed, where the operating procedures allowed to shift the chemical equilibrium of the reaction in the direction of the products of the process. Operations under hydrogen permeation, at 873 K, promoted the increase of methane conversion, circa 83%, and doubled the yield of hydrogen production, when compared with operations where no hydrogen permeation occurred. (author)

  10. Hydrogen production by reforming of liquid hydrocarbons in a membrane reactor for portable power generation-Experimental studies

    Science.gov (United States)

    Damle, Ashok S.

    One of the most promising technologies for lightweight, compact, portable power generation is proton exchange membrane (PEM) fuel cells. PEM fuel cells, however, require a source of pure hydrogen. Steam reforming of hydrocarbons in an integrated membrane reactor has potential to provide pure hydrogen in a compact system. Continuous separation of product hydrogen from the reforming gas mixture is expected to increase the yield of hydrogen significantly as predicted by model simulations. In the laboratory-scale experimental studies reported here steam reforming of liquid hydrocarbon fuels, butane, methanol and Clearlite ® was conducted to produce pure hydrogen in a single step membrane reformer using commercially available Pd-Ag foil membranes and reforming/WGS catalysts. All of the experimental results demonstrated increase in hydrocarbon conversion due to hydrogen separation when compared with the hydrocarbon conversion without any hydrogen separation. Increase in hydrogen recovery was also shown to result in corresponding increase in hydrocarbon conversion in these studies demonstrating the basic concept. The experiments also provided insight into the effect of individual variables such as pressure, temperature, gas space velocity, and steam to carbon ratio. Steam reforming of butane was found to be limited by reaction kinetics for the experimental conditions used: catalysts used, average gas space velocity, and the reactor characteristics of surface area to volume ratio. Steam reforming of methanol in the presence of only WGS catalyst on the other hand indicated that the membrane reactor performance was limited by membrane permeation, especially at lower temperatures and lower feed pressures due to slower reconstitution of CO and H 2 into methane thus maintaining high hydrogen partial pressures in the reacting gas mixture. The limited amount of data collected with steam reforming of Clearlite ® indicated very good match between theoretical predictions and

  11. Facility for Advanced Accelerator Experimental Tests (FACET) at SLAC and its Radiological Considerations

    Energy Technology Data Exchange (ETDEWEB)

    Mao, X.S.; Leitner, M.Santana; Vollaire, J.

    2011-08-22

    Facility for Advanced Accelerator Experimental Tests (FACET) in SLAC will be used to study plasma wakefield acceleration. FLUKA Monte Carlo code was used to design a maze wall to separate FACET project and LCLS project to allow persons working in FACET side during LCLS operation. Also FLUKA Monte Carlo code was used to design the shielding for FACET dump to get optimum design for shielding both prompt and residual doses, as well as reducing environmental impact. FACET will be an experimental facility that provides short, intense pulses of electrons and positrons to excite plasma wakefields and study a variety of critical issues associated with plasma wakefield acceleration [1]. This paper describes the FACET beam parameters, the lay-out and its radiological issues.

  12. Shoreline Erosion and Proposed Control at Experimental Facility 15-Spesutie Island

    Science.gov (United States)

    2017-09-01

    US Army Research Laboratory ATTN: RDRL-WMP-G Aberdeen Proving Ground , MD 21005-5066 8. PERFORMING ORGANIZATION REPORT NUMBER ARL-SR-0383 9...distribution is unlimited. 11 List of Symbols, Abbreviations, and Acronyms APG Aberdeen Proving Ground ARL US Army Research Laboratory EF 15...ARL-SR-0383 ● SEP 2017 US Army Research Laboratory Shoreline Erosion and Proposed Control at Experimental Facility 15–Spesutie

  13. Students' Assessment of Interactive Distance Experimentation in Nuclear Reactor Physics Laboratory Education

    Science.gov (United States)

    Malkawi, Salaheddin; Al-Araidah, Omar

    2013-01-01

    Laboratory experiments develop students' skills in dealing with laboratory instruments and physical processes with the objective of reinforcing the understanding of the investigated subject. In nuclear engineering, where research reactors play a vital role in the practical education of students, the high cost and long construction time of research…

  14. Experimental computer-controlled instrumentation system for the research reactor DR2

    DEFF Research Database (Denmark)

    Goodstein, L.P.

    1969-01-01

    An instrumentation system has been developed for one of the Danish Atomic Energy Commission's research reactors as part of an experiment on the advantages to be gained by the use of digital computers in a process plant application. Problem areas to be investigated include (a) reliability and safety...

  15. Experimental analysis of arsenic precipitation during microbial sulfate and iron reduction in model aquifer sediment reactors

    Science.gov (United States)

    Kirk, Matthew F.; Roden, Eric E.; Crossey, Laura J.; Brealey, Adrian J.; Spilde, Michael N.

    2010-05-01

    Microbial SO 42- reduction limits accumulation of aqueous As in reducing aquifers where the sulfide that is produced forms minerals that sequester As. We examined the potential for As partitioning into As- and Fe-sulfide minerals in anaerobic, semi-continuous flow bioreactors inoculated with 0.5% (g mL -1) fine-grained alluvial aquifer sediment. A fluid residence time of three weeks was maintained over a ca. 300-d incubation period by replacing one-third of the aqueous phase volume of the reactors with fresh medium every seven days. The medium had a composition comparable to natural As-contaminated groundwater with slightly basic pH (7.3) and 7.5 μM aqueous As(V) and also contained 0.8 mM acetate to stimulate microbial activity. Medium was delivered to a reactor system with and without 10 mmol L -1 synthetic goethite (α-FeOOH). In both reactors, influent As(V) was almost completely reduced to As(III). Pure As-sulfide minerals did not form in the Fe-limited reactor. Realgar (As 4S 4) and As 2S 3(am) were undersaturated throughout the experiment. Orpiment (As 2S 3) was saturated while sulfide content was low (˜50 to 150 μM), but precipitation was likely limited by slow kinetics. Reaction-path modeling suggests that, even if these minerals had formed, the dissolved As content of the reactor would have remained at hazardous levels. Mackinawite (Fe 1 + xS; x ⩽ 0.07) formed readily in the Fe-bearing reactor and held dissolved sulfide at levels below saturation for orpiment and realgar. The mackinawite sequestered little As (<0.1 wt.%), however, and aqueous As accumulated to levels above the influent concentration as microbial Fe(III) reduction consumed goethite and mobilized adsorbed As. A relatively small amount of pyrite (FeS 2) and greigite (Fe 3S 4) formed in the Fe-bearing reactor when we injected a polysulfide solution (Na 2S 4) to a final concentration of 0.5 mM after 216, 230, 279, and 286 days. The pyrite, and to a lesser extent the greigite, that formed

  16. COMPUTATIONAL AND EXPERIMENTAL MODELING OF THREE-PHASE SLURRY-BUBBLE COLUMN REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    Isaac K. Gamwo; Dimitri Gidaspow

    1999-09-01

    Considerable progress has been achieved in understanding three-phase reactors from the point of view of kinetic theory. In a paper in press for publication in Chemical Engineering Science (Wu and Gidaspow, 1999) we have obtained a complete numerical solution of bubble column reactors. In view of the complexity of the simulation a better understanding of the processes using simplified analytical solutions is required. Such analytical solutions are presented in the attached paper, Large Scale Oscillations or Gravity Waves in Risers and Bubbling Beds. This paper presents analytical solutions for bubbling frequencies and standing wave flow patterns. The flow patterns in operating slurry bubble column reactors are not optimum. They involve upflow in the center and downflow at the walls. It may be possible to control flow patterns by proper redistribution of heat exchangers in slurry bubble column reactors. We also believe that the catalyst size in operating slurry bubble column reactors is not optimum. To obtain an optimum size we are following up on the observation of George Cody of Exxon who reported a maximum granular temperature (random particle kinetic energy) for a particle size of 90 microns. The attached paper, Turbulence of Particles in a CFB and Slurry Bubble Columns Using Kinetic Theory, supports George Cody's observations. However, our explanation for the existence of the maximum in granular temperature differs from that proposed by George Cody. Further computer simulations and experiments involving measurements of granular temperature are needed to obtain a sound theoretical explanation for the possible existence of an optimum catalyst size.

  17. Theoretical and experimental study of the dark signal in CMOS image sensors affected by neutron radiation from a nuclear reactor

    Science.gov (United States)

    Xue, Yuanyuan; Wang, Zujun; He, Baoping; Yao, Zhibin; Liu, Minbo; Ma, Wuying; Sheng, Jiangkun; Dong, Guantao; Jin, Junshan

    2017-12-01

    The CMOS image sensors (CISs) are irradiated with neutron from a nuclear reactor. The dark signal in CISs affected by neutron radiation is studied theoretically and experimentally. The Primary knock-on atoms (PKA) energy spectra for 1 MeV incident neutrons are simulated by Geant4. And the theoretical models for the mean dark signal, dark signal non-uniformity (DSNU) and dark signal distribution versus neutron fluence are established. The results are found to be in good agreement with the experimental outputs. Finally, the dark signal in the CISs under the different neutron fluence conditions is estimated. This study provides the theoretical and experimental evidence for the displacement damage effects on the dark signal CISs.

  18. Theoretical and Experimental Evaluation of the Temperature Distribution in a Dry Type Air Core Smoothing Reactor of HVDC Station

    Directory of Open Access Journals (Sweden)

    Yu Wang

    2017-05-01

    Full Text Available The outdoor ultra-high voltage (UHV dry-type air-core smoothing reactors (DASR of High Voltage Direct Current systems are equipped with a rain cover and an acoustic enclosure. To study the convective heat transfer between the DASR and the surrounding air, this paper presents a coupled model of the temperature and fluid field based on the structural features and cooling manner. The resistive losses of encapsulations calculated by finite element method (FEM were used as heat sources in the thermal analysis. The steady fluid and thermal field of the 3-D reactor model were solved by the finite volume method (FVM, and the temperature distribution characteristics of the reactor were obtained. Subsequently, the axial and radial temperature distributions of encapsulation were investigated separately. Finally, an optical fiber temperature measurement scheme was used for an UHV DASR under natural convection conditions. Comparative analysis showed that the simulation results are in good agreement with the experimental data, which verifies the rationality and accuracy of the numerical calculation. These results can serve as a reference for the optimal design and maintenance of UHV DASRs.

  19. DIissolution of low enriched uranium from the experimental breeder reactor-II fuel stored at the Idaho National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Daniel, G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Almond, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); O' Rourke, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-28

    The Idaho National Laboratory (INL) is actively engaged in the development of electrochemical processing technology for the treatment of fast reactor fuels using irradiated fuel from the Experimental Breeder Reactor-II (EBR-II) as the primary test material. The research and development (R&D) activities generate a low enriched uranium (LEU) metal product from the electrorefining of the EBR-II fuel and the subsequent consolidation and removal of chloride salts by the cathode processor. The LEU metal ingots from past R&D activities are currently stored at INL awaiting disposition. One potential disposition pathway is the shipment of the ingots to the Savannah River Site (SRS) for dissolution in H-Canyon. Carbon steel cans containing the LEU metal would be loaded into reusable charging bundles in the H-Canyon Crane Maintenance Area and charged to the 6.4D or 6.1D dissolver. The LEU dissolution would be accomplished as the final charge in a dissolver batch (following the dissolution of multiple charges of spent nuclear fuel (SNF)). The solution would then be purified and the 235U enrichment downblended to allow use of the U in commercial reactor fuel. To support this potential disposition path, the Savannah River National Laboratory (SRNL) developed a dissolution flowsheet for the LEU using samples of the material received from INL.

  20. The National Ignition Facility: Status and Plans for the Experimental Program

    Energy Technology Data Exchange (ETDEWEB)

    Moses, E I

    2002-11-12

    The National Ignition Facility (NIF), currently under construction at the University of California's Lawrence Livermore National Laboratory, is a stadium-sized facility containing a 192-beam, 1.8-Megajoule, 500-Terawatt, 351-nm laser system and a 10-meter diameter target chamber with room for nearly 100 experimental diagnostics. NIF is being built by the National Nuclear Security Administration and when completed will be the world's largest laser experimental system, providing a national center to study inertial confinement fusion and the physics of matter at extreme energy densities and pressures. NIF's 192 energetic laser beams will compress fusion targets to conditions where they will ignite and burn, liberating more energy than required to initiate the fusion reactions. NIF experiments will allow the study of physical processes at temperatures approaching 100 million K and 100 billion times atmospheric pressure. These conditions exist naturally only in the interior of stars and in nuclear weapons explosions. In the course of designing the world's most energetic laser system, a number of significant technology breakthroughs have been achieved. NIF is now entering the first phases of its laser commissioning program. Low-energy preamplifier rod laser shots have been successfully propagated through the entire laser chain. Higher energy shots are planned through the end of 2002. NIF's target experimental systems are also being installed in preparation for laser performance and experimental capability commissioning starting in 2003.

  1. Automated operator procedure prompting for startup of Experimental Breeder Reactor-2

    Energy Technology Data Exchange (ETDEWEB)

    Renshaw, A.W.; Ball, S.J.; Ford, C.E.

    1990-11-01

    This report describes the development of an operator procedure prompting aid for startup of a nuclear reactor. This operator aid is a preliminary design for a similar aid that eventually will be used with the Advanced Liquid Metal Reactor (ALMR) presently in the design stage. Two approaches were used to develop this operator procedure prompting aid. One method uses an expert system software shell, and the other method uses database software. The preliminary requirements strongly pointed toward features traditionally associated with both database and expert systems software. Database software usually provides data manipulation flexibility and user interface tools, and expert systems tools offer sophisticated data representation and reasoning capabilities. Both methods, including software and associated hardware, are described in this report. Proposals for future enhancements to improve the expert system approach to procedure prompting and for developing other operator aids are also offered. 25 refs., 14 figs.

  2. A statistical experimental design to remove sulfate by crystallization in a fluidized-bed reactor

    OpenAIRE

    Mark Daniel G. de Luna; Rance, Diana Pearl M.; Luzvisminda M. Bellotindos; Lu, Ming-Chun

    2016-01-01

    This study used crystallization in a fluidized-bed reactor as an alternative technology to the conventional chemical precipitation to remove sulfate. The Box-Behnken Design was used to study the effects and interactions of seed dosage of synthetic gypsum, initial sulfate concentration and molar ratio of calcium to sulfate on conversion and removal of sulfate. The optimum conditions of conversion and removal of sulfate were determined and used to treat the simulated acid mine drainage (AMD) wa...

  3. Beam studies and experimental facility for the AWAKE experiment at CERN

    Energy Technology Data Exchange (ETDEWEB)

    Bracco, Chiara, E-mail: chiara.bracco@cern.ch; Gschwendtner, Edda, E-mail: edda.gschwendtner@cern.ch; Petrenko, Alexey, E-mail: alexey.petrenko@cern.ch; Timko, Helga, E-mail: helga.timko@cern.ch; Argyropoulos, Theodoros, E-mail: theodoros.argyropoulos@cern.ch; Bartosik, Hannes, E-mail: hannes.bartosik@cern.ch; Bohl, Thomas, E-mail: thomas.bohl@cern.ch; Esteban Müller, Juan, E-mail: juan.esteban.muller@cern.ch; Goddard, Brennan, E-mail: brennan.goddard@cern.ch; Meddahi, Malika, E-mail: malika.meddahi@cern.ch; Pardons, Ans, E-mail: ans.pardons@cern.ch; Shaposhnikova, Elena, E-mail: elena.chapochnikova@cern.ch; Velotti, Francesco M., E-mail: francesco.maria.velotti@cern.ch; Vincke, Helmut, E-mail: helmut.vincke@cern.ch

    2014-03-11

    A Proton Driven Plasma Wakefield Acceleration Experiment has been proposed as an approach to eventually accelerate an electron beam to the TeV energy range in a single plasma section. To verify this novel technique, a proof of principle R and D experiment, AWAKE, is planned at CERN using 400 GeV proton bunches from the SPS. An electron beam will be injected into the plasma cell to probe the accelerating wakefield. The AWAKE experiment will be installed in the CNGS facility profiting from existing infrastructure where only minor modifications need to be foreseen. The design of the experimental area and the proton and electron beam lines are shown. The achievable SPS proton bunch properties and their reproducibility have been measured and are presented. - Highlights: • A proton driven plasma wakefield experiment using the first time protons as drive beam is proposed. • The integration of AWAKE experiment, the proton, laser and electron beam line in an existing CERN facility is demonstrated. • The necessary modifications in the experimental facility are presented. • Proton beam optics and a new electron beam line are adapted to match with the required beam parameters. • Short high-intensity bunches were studied in the SPS to guide the design parameters of the AWAKE project.

  4. Stress state dependence of in-reactor creep and swelling. Part 2: Experimental results

    Science.gov (United States)

    Hall, M. M., Jr.; Flinn, J. E.

    2010-01-01

    Irradiation creep constitutive equations, which were developed in Part I, are used here to analyze in-reactor creep and swelling data obtained ca. 1977-1979 as part of the US breeder reactor program. The equations were developed according to the principles of incremental continuum plasticity for the purpose of analyzing data obtained from a novel irradiation experiment that was conducted, in part, using Type 304 stainless steel that had been previously irradiated to significant levels of void swelling. Analyses of these data support an earlier observation that all stress states, whether tensile, compressive, shear or mixed, can affect both void swelling and interactions between irradiation creep and swelling. The data were obtained using a set of five unique multiaxial creep-test specimens that were designed and used for the first time in this study. The data analyses demonstrate that the constitutive equations derived in Part I provide an excellent phenomenological representation of the interactive creep and swelling phenomena. These equations provide nuclear power reactor designers and analysts with a first-of-its-kind structural analysis tool for evaluating irradiation damage-dependent distortion of complex structural components having gradients in neutron damage rate, temperature and stress state.

  5. Experimental Studies on Assemblies 1 and 2 of the Fast Reactor FR-0. Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Hellstrand, E.; Andersson, T.L.; Brunfelter, B.; Kockum, J.; Londen, S.O.; Tiren, L.I.

    1965-12-15

    In a first part of this report, published as AE-195, an account was given of critical mass determinations and measurements of flux distribution and reaction ratios in the first assemblies of the fast zero power reactor FR0. This second part of the report deals with various investigations involving the measurement of reactivity. Control rod calibrations have been made using the positive period, the inverse multiplication, the rod drop and the pulsed source techniques, and show satisfactory agreement between the various methods. The reactivity worths of samples of different materials and different sizes have been measured at the core centre. Comparisons with perturbation calculations show that the regular and adjoint fluxes are well predicted in the central region of the core. The variation in the prompt neutron life-time with reactivity has been studied by means of the pulsed source and the Rossi-{alpha} techniques. Comparison with one region calculations reveals large discrepancies, indicating that this simple model is inadequate. Some investigations of streaming effects in an empty channel in the reactor and of interaction effects between channels have been made and are compared with theoretical estimates. Measurements of the reactivity worth of an air gap between the reactor halves and of the temperature coefficient are also described in the report. The work has been performed as a joint effort by AB Atomenergi and the Research Institute of National Defence.

  6. Proposal on experience learning of a nuclear reactor for children in future. A basic concept on a nuclear reactor facility for demonstration and education

    Energy Technology Data Exchange (ETDEWEB)

    Murata, Takashi [Kyoto Univ., Graduate School of Energy Science, Kyoto (Japan); Yoshiki, Nobuya; Kinehara, Yoshiki; Nakagawa, Haruo

    2001-12-01

    The Science Council of Japan indicates in a proposal on R and D on nuclear energy forward the 21st Century that it is important to expand the educational object on nuclear energy from colleges and gradual schools to elementary, middle high schools. And, the Committee of Japan Nuclear Energy Industries also proposed that as an effort forward security of reliability and popularization of knowledge, completeness of learning chance on energy and nuclear energy in education such as usage of general learning time, concept on establishment of educational reactor for demonstration and experience, is essential. Here was described on a concept on establishment of nuclear reactor for demonstration and experience at objectives of common national peoples, which was based on results of searches and investigations carried out by authors and aimed to supply to a field to grow up a literary adequately and widely capable of judging various information on the peoples by focusing to effectiveness of empirical learning as a method of promoting corrective understanding of common citizens on high class technical system and by establishment of the reactor aiming at general education on nuclear energy at a place easily accessible by common citizens, such as large city. (G.K.)

  7. Modeling and Experimental Studies of Mercury Oxidation and Adsorption in a Fixed-Bed and Entrained-Flow Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Buitrago, Paula A. [Univ. of Utah, Salt Lake City, UT (United States); Morrill, Mike [Univ. of Utah, Salt Lake City, UT (United States); Lighty, JoAnn S. [Univ. of Utah, Salt Lake City, UT (United States); Silcox, Geoffrey D. [Univ. of Utah, Salt Lake City, UT (United States)

    2009-06-01

    This report presents experimental and modeling mercury oxidation and adsorption data. Fixed-bed and single-particle models of mercury adsorption were developed. The experimental data were obtained with two reactors: a 300-W, methane-fired, tubular, quartz-lined reactor for studying homogeneous oxidation reactions and a fixed-bed reactor, also of quartz, for studying heterogeneous reactions. The latter was attached to the exit of the former to provide realistic combustion gases. The fixed-bed reactor contained one gram of coconut-shell carbon and remained at a temperature of 150°C. All methane, air, SO2, and halogen species were introduced through the burner to produce a radical pool representative of real combustion systems. A Tekran 2537A Analyzer coupled with a wet conditioning system provided speciated mercury concentrations. At 150°C and in the absence of HCl or HBr, the mercury uptake was about 20%. The addition of 50 ppm HCl caused complete capture of all elemental and oxidized mercury species. In the absence of halogens, SO2 increased the mercury adsorption efficiency to up to 30 percent. The extent of adsorption decreased with increasing SO2 concentration when halogens were present. Increasing the HCl concentration to 100 ppm lessened the effect of SO2. The fixed-bed model incorporates Langmuir adsorption kinetics and was developed to predict adsorption of elemental mercury and the effect of multiple flue gas components. This model neglects intraparticle diffusional resistances and is only applicable to pulverized carbon sorbents. It roughly describes experimental data from the literature. The current version includes the ability to account for competitive adsorption between mercury, SO2, and NO2. The single particle model simulates in-flight sorbent capture of elemental mercury. This model was developed to include Langmuir and Freundlich isotherms, rate equations, sorbent feed rate, and

  8. Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Yamaji, Bogdan; Aszodi, Attila [Budapest University of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-09-15

    In the paper measurement results from the experimental modelling of a molten salt reactor concept will be presented along with detailed uncertainty analysis of the experimental system. Non-intrusive flow measurements are carried out on the scaled and segmented mock-up of a homogeneous, single region molten salt fast reactor concept. Uncertainty assessment of the particle image velocimetry (PIV) measurement system applied with the scaled and segmented model is presented in detail. The analysis covers the error sources of the measurement system (laser, recording camera, etc.) and the specific conditions (de-warping of measurement planes) originating in the geometry of the investigated domain. Effect of sample size in the ensemble averaged PIV measurements is discussed as well. An additional two-loop-operation mode is also presented and the analysis of the measurement results confirm that without enhancement nominal and other operation conditions will lead to strong unfavourable separation in the core flow. It implies that use of internal flow distribution structures will be necessary for the optimisation of the core coolant flow. Preliminary CFD calculations are presented to help the design of a perforated plate located above the inlet region. The purpose of the perforated plate is to reduce recirculation near the cylindrical wall and enhance the uniformity of the core flow distribution.

  9. Potential for Ammonia Recapture by Farm Woodlands: Design and Application of a New Experimental Facility

    Directory of Open Access Journals (Sweden)

    Mark R. Theobald

    2001-01-01

    Full Text Available There has been increasing pressure on farmers in Europe to reduce the emissions of ammonia from their land. Due to the current financial climate in which farmers have to operate, it is important to identify ammonia control measures that can be adopted with minimum cost. The planting of trees around farmland and buildings has been identified as a potentially effective and low-cost measure to enhance ammonia recapture at a farm level and reduce long-range atmospheric transport. This work assesses experimentally what fraction of ammonia farm woodlands could potentially remove from the atmosphere. We constructed an experimental facility in southern Scotland to simulate a woodland shelterbelt planted in proximity to a small poultry unit. By measuring horizontal and vertical ammonia concentration profiles within the woodland, and comparing this to the concentration of an inert tracer (SF6 we estimate the depletion of ammonia due to dry deposition to the woodland canopy. Together with measurements of mean ammonia concentrations and throughfall fluxes of nitrogen, this information is used to provide a first estimate of the fraction of emitted ammonia that is recaptured by the woodland canopy. Analysis of these data give a lower limit of recapture of emitted ammonia, at the experimental facility, of 3%. By careful design of shelterbelt woodlands this figure could be significantly higher.

  10. Initial experimental results from the Laboratory Biosphere closed ecological system facility

    Science.gov (United States)

    Alling, A.; Allen, J.; Dempster, W.; Nelson, M.; Silverstone, S.; van Thillo, M.

    Results from the closure and initial closed ecological system research in the "Laboratory Biosphere" facility in Santa Fe, New Mexico (USA) will be presented. The facility was initially sealed in April 2002; and the first crop experiments with soybeans commenced in May 2002. The Laboratory Biosphere was created by the team which invented, built and operated Biosphere 2 during its years of closed ecological system functioning (1991-94) and is a testbed to build upon the lessons learned. It is an opportunity to continue experiments with a sustainable soil based agriculture system unlike most bioregenerative systems which use hydroponic systems dependent on a supply of nutrient solution. Because of the small volume of the system (34-45 m3), developing mechanisms to keep parameters like carbon dioxide within acceptable limits will be critical. Recycle of nutrients within the system to maintain soil fertility; and the ability of the inherent complex ecology of soils and a soil bed reactor to handle trace gas buildups are primary research goals. Other research goals are determination of short and long-term exchanges between soil, plants and atmosphere, especially for carbon dioxide, oxygen, nitrogen, NOX, and methane, impact of cultivation (tillage) on soil/atmospheric exchanges., investigation and development of strategies to return nutrients to the soil to maintain fertility, e.g. shredding biomass vs. composting, impact on soil chemistry of returning leachate water to the soil as irrigation water. The microbiological status of soils prior to experiments and over time will allow measurement of changes in microbial diversity and the determination of the role of soil microbes in biogeochemical cycles. Integration of automated sensor and control in the system with real-time modeling has importance for operation, research and educational outreach programs. The Laboratory Biosphere is intended to test and develop a "cybersphere" (network of shared intelligence) that may be

  11. Experimental investigation and model validation of a CaO/Ca(OH)2 fluidized bed reactor for thermochemical energy storage applications

    OpenAIRE

    Álvarez Criado, Yolanda; Huille, Alfred; Rougé, Sylvie; Abanades García, Juan Carlos

    2016-01-01

    The CaO/Ca(OH)2 hydration/dehydration chemical loop has long been recognized as a potential candidate for application in energy storage systems for concentrated solar plants. However, the technology still remains at a conceptual level because little information has been published on the performance of the key reactors in the system. In this work, we experimentally investigate the hydration and dehydration reactors in a 5.5 kWth batch fluidized bed reactor, in conditions relevant to larger sys...

  12. Instrumentation of VISTA test facility

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Seok; Choi, Ki Young; Park, Hyun Sik; Lee, Seong Jae; Park, Chun Kyong; Chung, Moon Ki

    2003-11-01

    VISTA (Experimental Verification by Integral Simulation of Transients and Accidents) is experimental facility to verify the performance and safety issues of SMART-P(Pilot plant of the System-integrated Modular Advanced Reactor), basic design of which has been completed by KAERI. The present report provide instrumentation details of VISTA in order to improve understanding on the phenomena and to certify the experimental data.

  13. Nuclear reactors built, being built, or planned, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Simpson, B.

    1992-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1991. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is an American company -- working either independently or in cooperation with a foreign company (Part 4, in each section). Critical assembly refers to an assembly of fuel and assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  14. Feasibility Study of a Novel Membrane Reactor for Syngas Production. Part 1: Experimental Study of O2 Permeation through Perovskite Membranes under Reducing and Non-Reducing Atmospheres

    NARCIS (Netherlands)

    Zhang Wenxing, Z.W.; Zhang, Wenxing; Smit, J.; van Sint Annaland, M.; Kuipers, J.A.M.

    2007-01-01

    In this contribution, the feasibility of a novel membrane reactor for energy efficient syngas production is investigated by means of an experimental and a simulation study. In Part 1, a detailed experimental study is performed on the O2 permeation through a perovskite membrane with composition

  15. Experimental subcritical facility driven by D-D/D-T neutron generator at BARC, India

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Amar, E-mail: image@barc.gov.in; Roy, Tushar; Kashyap, Yogesh; Ray, Nirmal; Shukla, Mayank; Patel, Tarun; Bajpai, Shefali; Sarkar, P.S.; Bishnoi, Saroj

    2015-05-01

    Highlights: •Experimental subcritical facility BRAHMMA coupled to D-D/D-T neutron generator. •Preliminary results of PNS experiments reported. •Feynman-alpha noise measurements explored with continuous source. -- Abstract: The paper presents design of an experimental subcritical assembly driven by D-D/D-T neutron and preliminary experimental measurements. The system has been developed for investigating the static and dynamic neutronic properties of accelerator driven sub-critical systems. This system is modular in design and it is first in the series of subcritical assemblies being designed. The subcritical core consists of natural uranium fuel with high density polyethylene as moderator and beryllium oxide as reflector. The fuel is embedded in high density polyethylene moderator matrix. Estimated k{sub eff} of the system is ∼0.89. One of the unique features of subcritical core is the use of Beryllium oxide (BeO) as reflector and HDPE as moderator making the assembly a compact modular system. The subcritical core is coupled to Purnima Neutron Generator which works in D-D and D-T mode with both DC and pulsed operation. It has facility for online source strength monitoring using neutron tagging and programmable source modulation. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques with D-D neutrons. Further experiments are being planned to measure the reactivity and other kinetic parameters using noise methods. This facility would also be used for carrying out studies on effect of source importance and measurement of source multiplication factor k{sub s} and external neutron source efficiency φ{sup ∗} in great details. Experiments with D-T neutrons are also underway.

  16. A facile and efficient method of enzyme immobilization on silica particles via Michael acceptor film coatings: immobilized catalase in a plug flow reactor.

    Science.gov (United States)

    Bayramoglu, Gulay; Arica, M Yakup; Genc, Aysenur; Ozalp, V Cengiz; Ince, Ahmet; Bicak, Niyazi

    2016-06-01

    A novel method was developed for facile immobilization of enzymes on silica surfaces. Herein, we describe a single-step strategy for generating of reactive double bonds capable of Michael addition on the surfaces of silica particles. This method was based on reactive thin film generation on the surfaces by heating of impregnated self-curable polymer, alpha-morpholine substituted poly(vinyl methyl ketone) p(VMK). The generated double bonds were demonstrated to be an efficient way for rapid incorporation of enzymes via Michael addition. Catalase was used as model enzyme in order to test the effect of immobilization methodology by the reactive film surface through Michael addition reaction. Finally, a plug flow type immobilized enzyme reactor was employed to estimate decomposition rate of hydrogen peroxide. The highly stable enzyme reactor could operate continuously for 120 h at 30 °C with only a loss of about 36 % of its initial activity.

  17. Operational readiness review for the Waste Experimental Reduction Facility. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1993-11-01

    An Operational Readiness Review (ORR) at the Idaho National Engineering Laboratory`s (INEL`s) Waste Experimental Reduction Facility (WERF) was conducted by EG&G Idaho, Inc., to verify the readiness of WERF to resume operations following a shutdown and modification period of more than two years. It is the conclusion of the ORR Team that, pending satisfactory resolution of all pre-startup findings, WERF has achieved readiness to resume unrestricted operations within the approved safety basis. ORR appraisal forms are included in this report.

  18. Integrated assessment of thermal hydraulic processes in W7-X fusion experimental facility

    Energy Technology Data Exchange (ETDEWEB)

    Kaliatka, T., E-mail: tadas.kaliatka@lei.lt; Uspuras, E.; Kaliatka, A.

    2017-02-15

    Highlights: • The model of Ingress of Coolant Event experiment facility was developed using the RELAP5 code. • Calculation results were compared with Ingress of Coolant Event experiment data. • Using gained experience, the numerical model of Wendelstein 7-X facility was developed. • Performed analysis approved pressure increase protection system for LOCA event. - Abstract: Energy received from the nuclear fusion reaction is one of the most promising options for generating large amounts of carbon-free energy in the future. However, physical and technical problems existing in this technology are complicated. Several experimental nuclear fusion devices around the world have already been constructed, and several are under construction. However, the processes in the cooling system of the in-vessel components, vacuum vessel and pressure increase protection system of nuclear fusion devices are not widely studied. The largest amount of radioactive materials is concentrated in the vacuum vessel of the fusion device. Vacuum vessel is designed for the vacuum conditions inside the vessel. Rupture of the in-vessel components of the cooling system pipe may lead to a sharp pressure increase and possible damage of the vacuum vessel. To prevent the overpressure, the pressure increase protection system should be designed and implemented. Therefore, systematic and detailed experimental and numerical studies, regarding the thermal-hydraulic processes in cooling system, vacuum vessel and pressure increase protection system, are important and relevant. In this article, the numerical investigation of thermal-hydraulic processes in cooling systems of in-vessel components, vacuum vessels and pressure increase protection system of fusion devices is presented. Using the experience gained from the modelling of “Ingress of Coolant Event” experimental facilities, the numerical model of Wendelstein 7-X (W7-X) experimental fusion device was developed. The integrated analysis of the

  19. Selective hydrogenation in trickle-bed reactor. Experimental and modelling including partial wetting.

    OpenAIRE

    Dietz, Adrian; Julcour-Lebigue, Carine; Wilhelm, Anne-Marie; Delmas, Henri

    2003-01-01

    International audience; A steady state model of a trickle bed reactor is developed for the consecutive hydrogenation of 1,5,9-cyclododecatriene on a Pd/Al2O3 catalyst. Various experiments have shown that the selectivity of this reaction towards the product of interest is much lower in co-current down-flow (trickle-bed) than in up-flow. This is due to uneven liquid distribution and to partial wetting of the catalyst surface at low liquid flow rates. The non-isothermal heterogeneous model propo...

  20. Characterization and quantification of an in-core neutron irradiation facility at a TRIGA II research reactor

    Science.gov (United States)

    Aghara, Sukesh; Charlton, William

    2006-07-01

    Experiments have been performed to characterize the neutron environment at an in-core TRIGA type nuclear research reactor. Steady-state thermal and epithermal neutron environment testing is important for many applications including, materials, electronics and biological cells. A well characterized neutron environment at a research reactor, including energy spectrum and spatial distribution, can be useful to many research communities and for educational research. This paper describes the characterization process and an application of exposing electronics to high neutron fluence.

  1. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor; Investigacao experimental da distribuicao de temperaturas no reator nuclear de pesquisa TRIGA IPR-R1

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Zacarias

    2005-07-01

    The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

  2. Experimental Development and Demonstration of Ultrasonic Measurement Diagnostics for Sodium Fast Reactor Thermal-hydraulics

    Energy Technology Data Exchange (ETDEWEB)

    Tokuhiro, Akira; Jones, Byron

    2013-09-13

    This research project will address some of the principal technology issues related to sodium-cooled fast reactors (SFR), primarily the development and demonstration of ultrasonic measurement diagnostics linked to effective thermal convective sensing under normatl and off-normal conditions. Sodium is well-suited as a heat transfer medium for the SFR. However, because it is chemically reactive and optically opaque, it presents engineering accessibility constraints relative to operations and maintenance (O&M) and in-service inspection (ISI) technologies that are currently used for light water reactors. Thus, there are limited sensing options for conducting thermohydraulic measurements under normal conditions and off-normal events (maintenance, unanticipated events). Acoustic methods, primarily ultrasonics, are a key measurement technology with applications in non-destructive testing, component imaging, thermometry, and velocimetry. THis project would have yielded a better quantitative and qualitative understanding of the thermohydraulic condition of solium under varied flow conditions. THe scope of work will evaluate and demonstrate ultrasonic technologies and define instrumentation options for the SFR.

  3. Experimental study of hydrodynamic and operation start of a baffled anaerobic reactor treating sewage

    Directory of Open Access Journals (Sweden)

    Ana Carolina Silveira Perico

    2009-12-01

    Full Text Available It is important to provide individual sanitation systems for sewage peri-urban communities or rural areas to minimize impacts on the environment and human health caused by sewage discharge in natura into water resources. In this context, the anaerobic digestion of effluent has been one of the main considered technologies due to easy implementation, material minimization and reduction in waste production. The objective of this work was to study a Baffled Anaerobic Reactor (BAR including its hydrodynamic characteristics, percentile of inoculum to be applied and reactor operation start. It was concluded that the flow is dispersed with 3.84% of dead spaces and that 20% of the cow manure provided best results; however, due to the high fiber content of the manure, its use is not recommended as inoculum. The BAR system, composed of four chambers, presented good performance for sewage treatment of a rural community in terms of organic substance removal (COD, turbidity and solids meeting effluent disposal standards of these parameters considering the Federal and Minas Gerais State legislation, in Brazil, even in a transient phase of operation, at temperatures below 20°C. However, the effluents from the BAR can’t be released into water bodies without other parameters such as nitrogen, phosphorus, fecal coliforms, and others are investigated to be conforming to those standards.

  4. Theoretical and experimental study of the photocatalytic activity of ZnO coated tubular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ríos-Valdovinos, E.; Amézaga-Madrid, P.; Antúnez-Flores, W.; Pola-Albores, F.; Pizá-Ruiz, P.; Miki-Yoshida, M., E-mail: mario.miki@cimav.edu.mx

    2014-01-15

    Highlights: • High quality ZnO thin films were deposited on the internal surface of fused silica tubing. • Surface carrier concentration was calculated theoretically under external irradiation. • Influence of film thickness on photocatalytic activity was explained by this model. • An optimum thickness around 60–70 nm was determined to get highest activity. -- Abstract: ZnO thin films were deposited inside of fused silica tubing by aerosol assisted chemical vapor deposition technique. The films were transparent, uniform, highly adherent and non-light scattering. Photocatalytic activity of internally ZnO coated tubing was evaluated by discoloration of a methyl orange aqueous solution in a batch reactor. Tubing was externally irradiated with UV-A at room temperature. A one dimensional model was proposed to calculate the spatial distribution of the carrier density and the films’ surface charge carrier concentration. This model can explain the influence of the films thickness on the photocatalytic activity. Results showed that the photocatalytic activity largely depends on the film thickness. For external irradiation of the films the optimum thickness was around 60–70 nm, for which the photocatalytic activity was maximum. The photonic efficiency of internally ZnO coated tubular reactors was evaluated as a function of initial colorant concentration, irradiation time and intensity. Furthermore, due to the high activity of the ZnO films, the films were repeatedly exposed to UV-A irradiation cycles, followed by activity measurement.

  5. Feasibility study for a biomedical experimental facility based on LEIR at CERN

    CERN Document Server

    Abler, Daniel; Carli, Christian; Dosanjh, Manjit; Peach, Ken; Orecchia, Roberto

    2013-01-01

    In light of the recent European developments in ion beam therapy, there is a strong interest from the biomedical research community to have more access to clinically relevant beams. Beamtime for pre-clinical studies is currently very limited and a new dedicated facility would allow extensive research into the radiobiological mechanisms of ion beam radiation and the development of more refined techniques of dosimetry and imaging. This basic research would support the current clinical efforts of the new treatment centres in Europe (for example HIT, CNAO and MedAustron). This paper presents first investigations on the feasibility of an experimental biomedical facility based on the CERN Low Energy Ion Ring LEIR accelerator. Such a new facility could provide beams of light ions (from protons to neon ions) in a collaborative and cost-effective way, since it would rely partly on CERN’s competences and infrastructure. The main technical challenges linked to the implementation of a slow extraction scheme for LEIR an...

  6. An Experimental Facility to Validate Ground Source Heat Pump Optimisation Models for the Australian Climate

    Directory of Open Access Journals (Sweden)

    Yuanshen Lu

    2017-01-01

    Full Text Available Ground source heat pumps (GSHPs are one of the most widespread forms of geothermal energy technology. They utilise the near-constant temperature of the ground below the frost line to achieve energy-efficiencies two or three times that of conventional air-conditioners, consequently allowing a significant offset in electricity demand for space heating and cooling. Relatively mature GSHP markets are established in Europe and North America. GSHP implementation in Australia, however, is limited, due to high capital price, uncertainties regarding optimum designs for the Australian climate, and limited consumer confidence in the technology. Existing GSHP design standards developed in the Northern Hemisphere are likely to lead to suboptimal performance in Australia where demand might be much more cooling-dominated. There is an urgent need to develop Australia’s own GSHP system optimisation principles on top of the industry standards to provide confidence to bring the GSHP market out of its infancy. To assist in this, the Queensland Geothermal Energy Centre of Excellence (QGECE has commissioned a fully instrumented GSHP experimental facility in Gatton, Australia, as a publically-accessible demonstration of the technology and a platform for systematic studies of GSHPs, including optimisation of design and operations. This paper presents a brief review on current GSHP use in Australia, the technical details of the Gatton GSHP facility, and an analysis on the observed cooling performance of this facility to date.

  7. Commissioning of the catalytic plasma exhaust clean-up facility caprice and first experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Glugla, M.; Kraemer, R.; Penzhorn, R.D.; Le, T.L.; Simon, K.H.; Guenther, K.; Besserer, U.; Schaefer, P.; Hellriegel, W. [Research Center Karlsruhe (Germany); Geissler, H. [Kraftanlagen Heidelberg (Germany)

    1995-10-01

    A fuel clean-up process for all plasma exhaust gases from DT fusion machines, based on catalytic conversion reactions combined with permeation of hydrogen isotopes through palladium/silver, has been developed. The complete process has already been proven with relevant concentrations of tritium at laboratory scale. On the basis of the results obtained the technical facility `CAPRICE` was designed, and is now under tritium operation at the Tritium Laboratory Karlsruhe (TLK). The facility is being used to demonstrate the process on a target throughput of 10 mol/h DT and 1 mol/h tritiated and non-tritiated impurities. Full scale experiments with hydrogen and deuterium have been completed to verify the design parameters of the facility and to gain detailed knowledge on the performance of the different subsystems under a variety of experimental conditions. Decontamination factors were obtained from these experiments as well as from first tritium runs employing about 350 Ci (0.5%) tritium in deuterium. 4 refs., 6 figs., 1 tab.

  8. MCNPCX calculations of dose rates and spectra in experimental channels of the CTEx irradiating facility

    Energy Technology Data Exchange (ETDEWEB)

    Gomes, Renato G.; Rebello, Wilson F.; Vellozo, Sergio O.; Junior, Luis M., E-mail: renatoguedes@ime.eb.br, E-mail: rebello@ime.eb.br, E-mail: vellozo@cbpf.br, E-mail: luisjrmoreira@hotmail.com [Instituto Militar de Engenharia (IME), Janeiro, RJ (Brazil); Vital, Helio C., E-mail: vital@ctex.eb.br [Centro Tecnologico do Exercito (CTEx), Barra de Guaratiba, RJ (Brazil); Rusin, Tiago, E-mail: tiago.rusin@mma.gov.br [Ministerio do Meio Ambiente, Brasilia, DF (Brazil); Silva, Ademir X., E-mail: ademir@con.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    MCNPX simulations have been performed in order to calculate dose rates as well as spectra along the four experimental channels of the gamma irradiating facility at the Technology Center of the Brazilian Army (CTEx). Safety, operational and research requirements have led to the need to determine both the magnitude and spectra of the leaking gamma fluxes. The CTEx experimental facility is cavity type with a moveable set of 28 horizontally positioned rods, filled with Cesium-137 chloride and doubly encased in stainless steel that yields an approximately plane 42 kCi-source that provides a maximum dose rate of about 1.5 kG/h into two irradiating chambers. The channels are intended for irradiation tests outside facility. They would allow larger samples to be exposed to lower gamma dose rates under controlled conditions. Dose rates have been calculated for several positions inside the channels as well as at their exits. In addition, for purposes related to the safety of operators and personnel, the angles submitted by the exiting beams have also been evaluated as they spread when leaving the channels. All calculations have been performed by using a computational model of the CTEx facility that allows its characteristics and operation to be accurately simulated by using the Monte Carlo Method. Virtual dosimeters filled with Fricke (ferrous sulfate) were modeled and positioned throughout 2 vertical channels (top and bottom) and 2 horizontal ones (front and back) in order to map dose rates and gamma spectrum distributions. The calculations revealed exiting collimated beams in the order of tenths of Grays per minute as compared to the maximum 25 Gy / min dose rate in the irradiator chamber. In addition, the beams leaving the two vertical channels were found to exhibit a widespread cone-shaped distribution with aperture angle ranging around 85 deg. The data calculated in this work are intended for use in the design of optimized experiments (better positioning of samples and

  9. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  10. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1994. Twenty-seventh annual report

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, M.L.; Hagemeyer, D. [Science Applications International Corporation, Oak Ridge, TN (United States)

    1996-01-01

    This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission`s (NRC) Radiation Exposure Information and Reporting System (REIRS). Annual reports for 1994 were received from a total of 303 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 303 licensees indicated that 152,028 individuals were monitored, 79,780 of whom received a measurable dose. The collective dose incurred by these individuals was 24,740 person-cSv (person-rem){sup 2} which represents a 15% decrease from the 1993 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.31 cSv (rem) for 1994. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. In 1994, the annual collective dose per reactor for light water reactor licensees (LWRs) was 198 person-cSv (person-rem). This represents a 18% decrease from the 1993 value of 242 person-cSv (person-rem). The annual collective dose per reactor for boiling water reactors (BWRs) was 327 person-cSv (person-rem) and, for pressurized water reactors (PWRs), it was 131 person-cSv (person-rem). Analyses of transient worker data indicate that 18,178 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1994, the average measurable dose calculated from reported data was 0.28 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.31 cSv (rem).

  11. The experimental facility for investigation of MHD heat transfer in perspective coolants in nuclear energetics.

    Science.gov (United States)

    Batenin, B. M.; Belyaev, I. A.; Birukov, D. A.; Frick, P. G.; Nikitina, I. S.; Manchkha, S. P.; Pyatnitskaya, N. Yu; Razuvanov, N. G.; Sviridov, E. V.; Sviridov, V. G.

    2017-11-01

    Paper presents the current results of work conducted by a joint research group of MPEI–JIHT RAS for experimental study of liquid metals heat transfer. The team of specialists of MPEI–JIHT RAS put into operation a new mercury MHD facility RK-3. The main components of this stand are: a unique electromagnet, created by specialists of the Budker Institute of Nuclear Physics (BINP), and a sealed liquid-metal circuit. The facility will be explored lifting and standpipe flow of liquid metal in a transverse magnetic field in channels of different forms. For the experiments on the study of heat transfer and hydrodynamics of flows for measuring characteristics such as temperature, speed, pulse characteristics, probe method is used. Presents the first experimental results obtained for a pipe in a transverse magnetic field. During the experiments with various flow parameters data was obtained and processed with constructing temperature fields, dimensionless wall temperature distributions and heat transfer coefficients along the perimeter of the work area. Modes with low frequency pulsations of temperature were discovered. The boundaries where low frequency temperature fluctuations occur were defined in a circular tube.

  12. SIRHEX—A new experimental facility for high heat flux testing of plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Kunze, André, E-mail: andre.kunze@kit.edu [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (Germany); Ghidersa, Bradut-Eugen [Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology (Germany); Bonelli, Flavia [Politecnico di Torino, Dipartimento Energia (Italy)

    2015-10-15

    Highlights: • Commercial infrared heaters have been qualified for future First Wall experiments. • In first tests surface heat flux densities up to 470 kW/m were achieved. • The homogeneity of the heat distribution stayed within ±5% of the nominal value. • With the heaters a typical ITER pulse can be reproduced. • An adequate testing strategy will be required to improve heater lifetime. - Abstract: SIRHEX (“Surface Infrared Radiation Heating Experiment”) is a small-scale experimental facility at KIT, which has been built for testing and qualifying high heat flux radiation heaters for blanket specific conditions using an instrumented water cooled target. This paper describes the SIRHEX facility and the experimental set-up for the heater tests. The results of a series of tests focused on reproducing homogeneous surface heat flux densities up to 500 kW/m{sup 2} will be presented and the impact of the heater performance on the design of the First Wall test rig will be discussed.

  13. Experimental subcritical facility driven by D-D/D-T neutron generator at BARC, India

    Science.gov (United States)

    Sinha, Amar; Roy, Tushar; Kashyap, Yogesh; Ray, Nirmal; Shukla, Mayank; Patel, Tarun; Bajpai, Shefali; Sarkar, P. S.; Bishnoi, Saroj

    2015-05-01

    The paper presents design of an experimental subcritical assembly driven by D-D/D-T neutron and preliminary experimental measurements. The system has been developed for investigating the static and dynamic neutronic properties of accelerator driven sub-critical systems. This system is modular in design and it is first in the series of subcritical assemblies being designed. The subcritical core consists of natural uranium fuel with high density polyethylene as moderator and beryllium oxide as reflector. The fuel is embedded in high density polyethylene moderator matrix. Estimated keff of the system is ∼0.89. One of the unique features of subcritical core is the use of Beryllium oxide (BeO) as reflector and HDPE as moderator making the assembly a compact modular system. The subcritical core is coupled to Purnima Neutron Generator which works in D-D and D-T mode with both DC and pulsed operation. It has facility for online source strength monitoring using neutron tagging and programmable source modulation. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques with D-D neutrons. Further experiments are being planned to measure the reactivity and other kinetic parameters using noise methods. This facility would also be used for carrying out studies on effect of source importance and measurement of source multiplication factor ks and external neutron source efficiency φ∗ in great details. Experiments with D-T neutrons are also underway.

  14. Facility for Advanced Accelerator Experimental Tests at SLAC (FACET) Conceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Amann, J.; Bane, K.; /SLAC

    2009-10-30

    This Conceptual Design Report (CDR) describes the design of FACET. It will be updated to stay current with the developing design of the facility. This CDR begins as the baseline conceptual design and will evolve into an 'as-built' manual for the completed facility. The Executive Summary, Chapter 1, gives an introduction to the FACET project and describes the salient features of its design. Chapter 2 gives an overview of FACET. It describes the general parameters of the machine and the basic approaches to implementation. The FACET project does not include the implementation of specific scientific experiments either for plasma wake-field acceleration for other applications. Nonetheless, enough work has been done to define potential experiments to assure that the facility can meet the requirements of the experimental community. Chapter 3, Scientific Case, describes the planned plasma wakefield and other experiments. Chapter 4, Technical Description of FACET, describes the parameters and design of all technical systems of FACET. FACET uses the first two thirds of the existing SLAC linac to accelerate the beam to about 20GeV, and compress it with the aid of two chicanes, located in Sector 10 and Sector 20. The Sector 20 area will include a focusing system, the generic experimental area and the beam dump. Chapter 5, Management of Scientific Program, describes the management of the scientific program at FACET. Chapter 6, Environment, Safety and Health and Quality Assurance, describes the existing programs at SLAC and their application to the FACET project. It includes a preliminary analysis of safety hazards and the planned mitigation. Chapter 7, Work Breakdown Structure, describes the structure used for developing the cost estimates, which will also be used to manage the project. The chapter defines the scope of work of each element down to level 3.

  15. Fluid flow separation in a reactor pressure vessel during an ECC injection. Single phase flow and two phase flow (air-water) experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Thierry Bichet; Alain Martin [EDF - Research and Development Division - Fluid Mechanics and Heat Transfert 6, quai Watier - B.P. 49 - 78401 Chatou CEDEX 01 (France); Frederic Beaud [EDF/ Industry - Basic Design Department., 12-14, Avenue Dutrievoz 69628 Villeurbanne CEDEX (France)

    2005-07-01

    Full text of publication follows: Within the framework of the nuclear power plant lifetime issue, the assessment of the French 900 MWe (3-loops) series reactor pressure vessel (RPV) integrity has been performed. A simplified analysis has shown that the most severe loading conditions are given by the small break loss of coolant accidents due to the pressurized injection of cold water (9 deg. C) into the cold leg and down comer of the RPV. During these transient scenarios, single or two-phase (uncovered cold leg) flows have been shown in the cold leg, depending on the crack size and RPV model (900 MWe or 1300 MWe). An experimental study has been carried out, on the one hand, to consolidate the numerical results obtained with CFD home code (Code-Saturne) which mainly showed the stratified flow in the cold leg and the fluid flow separation and its oscillations in the down comer during a single phase scenario. These physical phenomena are important for the thermal RPV loading assessment. On the other hand, the absence of experimental two-phase data necessitated to carry out an experimental study around the mixing area behavior (free surface, stratified flow) during an ECC injection with an uncovered cold leg. The new EDF R and D mock up, called HYBISCUS, is a facility which is made out of Plexiglas (atmosphere pressure) and represents a half scale CP0 geometry with one cold leg and part of the down comer. The mock up modularity allows us to insert representative ECC nozzles and a thermal shield. In reference to the reactor scenarios, the experimental operating conditions are derived from the conservation of the density effects (Froude number). For that, a heated salted water flow is used to represent the ECC injection whereas water represents the cold leg fluid. This mock up has been defined in order to represent single phase flow (cold leg and down comer full of water) or two-phase flow (uncovered cold leg) ECC scenarios. This paper reports experimental results

  16. Experimental assessment of accident scenarios for the high temperature reactor fuel system

    Energy Technology Data Exchange (ETDEWEB)

    Seeger, O.; Laurie, M.; Bottomley, P.D.W.; Rondinella, V.V. [European Commission, Joint Research Center, Karlsruhe (Germany). Inst. for Transuranium Elements; Avincola, V. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany). Inst. fuer Angewandte Materialien (IAM-AWP); Allelein, H.J. [RWTH Aachen Univ. (Germany). Lehrstuhl fuer Reaktorsicherheit und -technik

    2013-11-15

    The High Temperature Reactor (HTR) is characterized by an advanced design with passive safety features. Fuel elements are constituted by a graphite matrix containing sub-mm-sized fuel particles with TRi-ISOtropic (TRISO) coating, designed to provide high fission product retention. During a loss of coolant accident scenario in a HTR the maximum temperature is foreseen to be in the range of 1,600 to 1,650 C, remaining well below the melting point of the fuel. Two key aspects associated with the safety of HTR fuel are assessed in this paper: fission product retention at temperatures up to 1,800 C is analyzed with the Cold Finger Apparatus (KueFA) while the behaviour of HTR-relevant fuel materials in an oxidizing environment is studied with the Corrosion Apparatus KORA. The KueFA is used to observe the combined effects of Depressurization and LOss of Forced Circulation (DLOFC) accident scenarios on HTR fuel. Originally designed at the Forschungszentrum Juelich (FZJ), an adapted KueFA operates on irradiated fuel in hot cell at JRC-ITU. A fuel pebble is heated in helium atmosphere for several hundred hours, mimicking accident temperatures up to 1,800 C and realistic temperature transients. Nongaseous volatile fission products released from the fuel condense on a water cooled stainless steel plate dubbed 'Cold Finger'. Exchanging plates frequently during the experiment and analyzing plate deposits by means of High Purity Germanium (HPGe) gamma spectroscopy allows a reconstruction of the fission product release as a function of time and temperature. To achieve a good quantification of the release, a careful calibration of the setup is necessary and a collimator needs to be used in some cases. The analysis of condensation plates from recent KueFA tests shows that fission product release quantification is possible at high and low activity levels. Another relevant HTR accident scenario is air ingress into the reactor vessel as a consequence of a DLOFC incident. In

  17. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1995: Twenty-eighth annual report. Volume 17

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, M.L. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications; Hagemeyer, D. [Science Applications International Corp., Oak Ridge, TN (United States)

    1997-01-01

    This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission`s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1995 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high-level waste currently licensed, only six categories will be considered in this report. In 1995, the annual collective dose per reactor for light water reactor licensees (LWRs) was 199 person-cSv (person-rem). This is the same value that was reported for 1994. The annual collective dose per reactor for boiling water reactors (BWRs) was 256 person-cSv (person-rem) and, for pressurized water reactors (PWRs), it was 170 person-cSv (person-rem). Analyses of transient worker data indicate that 17,153 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1995, the average measurable dose calculated from reported data was 0.26 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.32 cSv (rem).

  18. Experimental investigation of the solar carbothermic reduction of ZnO using a two-cavity solar reactor

    Energy Technology Data Exchange (ETDEWEB)

    Frommherz, U.; Osinga, T.; Steinfeld, A.; Wieckert, C.

    2003-03-01

    Zinc production by solar carbothermic reduction of ZnO offers a CO{sub 2} emission reduction by a factor of 5 vis-a-vis the conventional fossil-fuel-based electrolytic or Imperial Smelting processes. Zinc can serve as a fuel in Zn-air fuel cells or can be further reacted with H{sub 2}O to form high-purity H{sub 2}. In either case, the product ZnO can be solar-recycled to Zn. We report on experimental results obtained with a 5 kW solar chemical reactor prototype that features two cavities in series, with the inner one functioning as the solar absorber and the outer one as the reaction chamber. Tests were conducted at PSI's Solar Furnace and ETH's High-Flux Solar Simulator to investigate the effect of process temperature (range 1350-1600 K) and reducing agent type (beech charcoal, activated charcoal, petcoke) on the reactor's performance and on the chemical conversion. In a typical 40-min solar experiment at 1500 K, 500 g of a 1:0.8 stoichiometric ZnO-C mixture were processed into Zn(g), CO, and CO{sub 2}. Thermal efficiencies of up to 20% were achieved. (author)

  19. Experimental Design for Evaluating Selected Nondestructive Measurement Technologies - Advanced Reactor Technology Milestone: M3AT-16PN2301043

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, Evelyn H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pitman, Stan G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dib, Gerges [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Roy, Surajit [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Good, Morris S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Walker, Cody M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-07-16

    The harsh environments in advanced reactors (AdvRx) increase the possibility of degradation of safety-critical passive components, and therefore pose a particular challenge for deployment and extended operation of these concepts. Nondestructive evaluation technologies are an essential element for obtaining information on passive component condition in AdvRx, with the development of sensor technologies for nondestructively inspecting AdvRx passive components identified as a key need. Given the challenges posed by AdvRx environments and the potential needs for reducing the burden posed by periodic in-service inspection of hard-to-access and hard-to-replace components, a viable solution may be provided by online condition monitoring of components. This report identifies the key challenges that will need to be overcome for sensor development in this context, and documents an experimental plan for sensor development, test, and evaluation. The focus of initial research and development is on sodium fast reactors, with the eventual goal of the research being developing the necessary sensor technology, quantifying sensor survivability and long-term measurement reliability for nondestructively inspecting critical components. Materials for sensor development that are likely to withstand the harsh environments are described, along with a status on the fabrication of reference specimens, and the planned approach for design and evaluation of the sensor and measurement technology.

  20. Combined numerical and experimental investigation into the coolant flow hydrodynamics and mass transfer behind the spacer grid in fuel assemblies of the floating power unit reactor

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-09-01

    Full Text Available The results of experimental investigations into the local hydrodynamics and inter-cell mass transfer of the coolant flow in representative zones of the KLT-40C reactor FAs behind the plate-type spacer grid are presented. The investigations were conducted on an aerodynamic rig using the admixture diffusion method (the tracer-gas method. A study into the spatial dispersion of the absolute flow velocity projections and into the distribution of the tracer concentration allowed specify the coolant flow pattern behind the FA plate-type spacer grid of the KLT-40C reactor. The results of measuring the flow friction coefficient in the plate-type spacer grid, depending on the Reynolds number, are presented. Based on the obtained experimental data, recommendations have been provided for updating the procedures to calculate the coolant flow rates for the KLT-40C reactor core by-channel codes. The results of investigating the coolant flow local hydrodynamics and mass transfer in the KLT-40C reactor FAs have been adopted for practical use by Afrikantov OKBM for estimating the heat-engineering reliability of the KLT-40C reactor cores and have been data based for verification of CFD codes and detailed by-channel calculation of the KLT-40C reactor core.

  1. The experimental set-up of the RIB in-flight facility EXOTIC

    Science.gov (United States)

    Pierroutsakou, D.; Boiano, A.; Boiano, C.; Di Meo, P.; La Commara, M.; Manea, C.; Mazzocco, M.; Nicoletto, M.; Parascandolo, C.; Signorini, C.; Soramel, F.; Strano, E.; Toniolo, N.; Torresi, D.; Tortone, G.; Anastasio, A.; Bettini, M.; Cassese, C.; Castellani, L.; Corti, D.; Costa, L.; De Fazio, B.; Galet, G.; Glodariu, T.; Grebosz, J.; Guglielmetti, A.; Molini, P.; Pontoriere, G.; Rocco, R.; Romoli, M.; Roscilli, L.; Sandoli, M.; Stroe, L.; Tessaro, M.; Zatti, P. G.

    2016-10-01

    We describe the experimental set-up of the Radioactive Ion Beam (RIB) in-flight facility EXOTIC consisting of: (a) two position-sensitive Parallel Plate Avalanche Counters (PPACs), dedicated to the event-by-event tracking of the produced RIBs and to time of flight measurements and (b) the new high-granularity compact telescope array EXPADES (EXotic PArticle DEtection System), designed for nuclear physics and nuclear astrophysics experiments employing low-energy light RIBs. EXPADES consists of eight ΔE -Eres telescopes arranged in a cylindrical configuration around the target. Each telescope is made up of two Double Sided Silicon Strip Detectors (DSSSDs) with a thickness of 40/60 μm and 300 μm for the ΔE and Eres layer, respectively. Additionally, eight ionization chambers were constructed to be used as an alternative ΔE stage or, in conjunction with the entire DSSSD array, to build up more complex triple telescopes. New low-noise multi-channel charge-sensitive preamplifiers and spectroscopy amplifiers, associated with constant fraction discriminators, peak-and-hold and Time to Amplitude Converter circuits were developed for the electronic readout of the ΔE stage. Application Specific Integrated Circuit-based electronics was employed for the treatment of the Eres signals. An 8-channel, 12-bit multi-sampling 50 MHz Analog to Digital Converter, a Trigger Supervisor Board for handling the trigger signals of the whole experimental set-up and an ad hoc data acquisition system were also developed. The performance of the PPACs, EXPADES and of the associated electronics was obtained offline with standard α calibration sources and in-beam by measuring the scattering process for the systems 17O+58Ni and 17O+208Pb at incident energies around their respective Coulomb barriers and, successively, during the first experimental runs with the RIBs of the EXOTIC facility.

  2. Stabilized Spheromak Fusion Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  3. Experimental Investigations of Physical and Chemical Properties for Microalgae HTL Bio-Crude Using a Large Batch Reactor

    Directory of Open Access Journals (Sweden)

    Farhad M. Hossain

    2017-04-01

    Full Text Available As a biofuel feedstock, microalgae has good scalability and potential to supply a significant proportion of world energy compared to most types of biofuel feedstock. Hydrothermal liquefaction (HTL is well-suited to wet biomass (such as microalgae as it greatly reduces the energy requirements associated with dewatering and drying. This article presents experimental analyses of chemical and physical properties of bio-crude oil produced via HTL using a high growth-rate microalga Scenedesmus sp. in a large batch reactor. The overarching goal was to investigate the suitability of microalgae HTL bio-crude produced in a large batch reactor for direct application in marine diesel engines. To this end we characterized the chemical and physical properties of the bio-crudes produced. HTL literature mostly reports work using very small batch reactors which are preferred by researchers, so there are few experimental and parametric measurements for bio-crude physical properties, such as viscosity and density. In the course of this study, a difference between traditionally calculated values and measured values was noted. In the parametric study, the bio-crude viscosity was significantly closer to regular diesel and biodiesel standards than transesterified (FAME microalgae biodiesel. Under optimised conditions, HTL bio-crude’s high density (0.97–1.04 kg·L−1 and its high viscosity (70.77–73.89 mm2·s−1 had enough similarity to marine heavy fuels. although the measured higher heating value, HHV, was lower (29.8 MJ·kg−1. The reaction temperature was explored in the range 280–350 °C and bio-crude oil yield and HHV reached their maxima at the highest temperature. Slurry concentration was explored between 15% and 30% at this temperature and the best HHV, O:C, and N:C were found to occur at 25%. Two solvents (dichloromethane and n-hexane were used to recover the bio-crude oil, affecting the yield and chemical composition of the bio-crude.

  4. Experimental investigation of pebble flow dynamics using radioactive particle tracking technique in a scaled-down Pebble Bed Modular Reactor (PBMR)

    Energy Technology Data Exchange (ETDEWEB)

    Khane, Vaibhav; Said, I.A.; Al-Dahhan, Muthanna H., E-mail: aldahhanm@mst.edu

    2016-06-15

    Highlights: • Pebble Flow fields at Pebble Bed Modular Reactor was investigated. • Radioactive Particle Tracking (RPT) technique has been used. • Plug flow type velocity profile is suggested at upper cylindrical region. - Abstract: The Pebble Bed Modular Reactor (PBMR) is a type of very-high-temperature reactor (VHTR) that is conceptually very similar to moving bed reactors used in the chemical and petrochemical industries. In a PBMR core, nuclear fuel is in the form of pebbles and moves slowly under the influence of gravity. In this work, an integrated experimental and computational study of granular flow in a scaled-down cold flow PBMR was performed. A continuous pebble re-circulation experimental set-up, mimicking the flow of pebbles in a PBMR was designed and developed. An experimental investigation of pebble flow dynamics in a scaled down test reactor was carried out using a non-invasive radioactive particle tracking (RPT) technique that used a cobalt-60 based tracer to mimic pebbles in terms of shape, size and density. A cross-correlation based position reconstruction algorithm and RPT calibration data were used to obtain results about Lagrangian trajectories, the velocity field, and residence time distributions. The RPT technique results a serve as a benchmark data for assessing contact force models used in the discrete element method (DEM) simulations.

  5. The New Cold Neutron Radiography Facility (CNRF) at the Mianyang Research Reactor of the China Academy of Engineering Physics

    Science.gov (United States)

    Bin, Tang; Heyong, Huo; Ke, Tang; Rogers, John; Haste, Martin; Christodoulou, Marios

    A new cold neutron radiography beamline has been designed and constructed for the Mianyang reactor at the Institute of Nuclear Physics and Chemistry of the China Academy of Engineering Physics. This paper describes the components of the system and demonstrates the achievable image resolution.

  6. Numerical and experimental investigation of surface vortex formation in coolant reservoirs of reactor safety systems

    Energy Technology Data Exchange (ETDEWEB)

    Pandazis, Peter [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Babcsany, Boglarka [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques

    2016-11-15

    The reliable operation of the emergency coolant pumps and passive gravitational injection systems are an important safety issue during accident scenarios with coolant loss in pressurized water reactors. Because of the pressure drop and flow disturbances surface vortices develops at the pump intakes if the water level decreasing below a critical value. The induced swirling flow and gas entrainment lead to flow limitation and to pump failures and damages. The prediction of the critical submergence to avoid surface vortex building is difficult because it depends on many geometrical and fluid dynamical parameters. An alternative and new method has been developed for the investigation of surface vortices. The method based on the combination of CFD results with the analytical vortex model of Burgers and Rott. For further investigation the small scale experiments from the Institute of Nuclear Techniques of the Budapest University of Technology and Economics are used which were inspired from flow limitation problems during the draining of the bubble condenser trays at a VVER type nuclear power plants.

  7. The thermal decomposition of the benzyl radical in a heated micro-reactor. I. Experimental findings

    Science.gov (United States)

    Buckingham, Grant T.; Ormond, Thomas K.; Porterfield, Jessica P.; Hemberger, Patrick; Kostko, Oleg; Ahmed, Musahid; Robichaud, David J.; Nimlos, Mark R.; Daily, John W.; Ellison, G. Barney

    2015-01-01

    The pyrolysis of the benzyl radical has been studied in a set of heated micro-reactors. A combination of photoionization mass spectrometry (PIMS) and matrix isolation infrared (IR) spectroscopy has been used to identify the decomposition products. Both benzyl bromide and ethyl benzene have been used as precursors of the parent species, C6H5CH2, as well as a set of isotopically labeled radicals: C6H5CD2, C6D5CH2, and C6H513CH2. The combination of PIMS and IR spectroscopy has been used to identify the earliest pyrolysis products from benzyl radical as: C5H4=C=CH2, H atom, C5H4—C ≡ CH, C5H5, HCCCH2, and HC ≡ CH. Pyrolysis of the C6H5CD2, C6D5CH2, and C6H513CH2 benzyl radicals produces a set of methyl radicals, cyclopentadienyl radicals, and benzynes that are not predicted by a fulvenallene pathway. Explicit PIMS searches for the cycloheptatrienyl radical were unsuccessful, there is no evidence for the isomerization of benzyl and cycloheptatrienyl radicals: C6H5CH2⇋C7H7. These labeling studies suggest that there must be other thermal decomposition routes for the C6H5CH2 radical that differ from the fulvenallene pathway.

  8. Experimental Facility for Checking the Possibility to Obtain Super-High Temperature Due to Acoustic Cavitation

    CERN Document Server

    Miller, M B; Sobolev, Yu G; Kostenko, B F

    2004-01-01

    An experimental facility developed for checking the possibility to obtain super-high temperature sufficient for thermonuclear reaction D($d, n$)$^{3}$He in an acoustic cavitation is described. The acoustic part of the instrumentation consists of a resonator and a system exciting high amplitude of the acoustic field within the resonator. The cavitation process is controlled with the use of fast neutron pulses. The instrument includes a system of pumping out solute gases from the liquid (acetone enriched with deuterium up to 99{\\%}) without losses of matter. Measuring of the field is based on the calibration procedure including observation of sonoluminescence. The system of detection and identification of D($d, n$)$^{3}$He reaction is based on a scintillation detector of fast neutrons and a system of measuring multiparameter events by the correlation technique with separation of the neutrons from the $\\gamma $-radiation background (pulse shape discrimination).

  9. Survey of existing underground openings for in-situ experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Wollenberg, H.; Graf, A.; Strisower, B.; Korbin, G.

    1981-07-01

    In an earlier project, a literature search identified 60 underground openings in crystalline rock capable of providing access for an in-situ experimental facility to develop geochemical and hydrological techniques for evaluating sites for radioactive waste isolation. As part of the current project, discussions with state geologists, owners, and operators narrowed the original group to 14. Three additional sites in volcanic rock and one site in granite were also identified. Site visits and application of technical criteria, including the geologic and hydrologic settings and depth, extent of the rock unit, condition, and accessibility of underground workings, determined four primary candidate sites: the Helms Pumped Storage Project in grandiodorite of the Sierra Nevada, California; the Tungsten Queen Mine in Precambrian granodiorite of the North Carolina Piedmont; the Mount Hope Mine in Precambrian granite and gneiss of northern New Jersey; and the Minnamax Project in the Duluth gabbro complex of northern Minnesota.

  10. NASA HRP Plans for Collaboration at the IBMP Ground-Based Experimental Facility (NEK)

    Science.gov (United States)

    Cromwell, Ronita L.

    2016-01-01

    NASA and IBMP are planning research collaborations using the IBMP Ground-based Experimental Facility (NEK). The NEK offers unique capabilities to study the effects of isolation on behavioral health and performance as it relates to spaceflight. The NEK is comprised of multiple interconnected modules that range in size from 50-250m(sup3). Modules can be included or excluded in a given mission allowing for flexibility of platform design. The NEK complex includes a Mission Control Center for communications and monitoring of crew members. In an effort to begin these collaborations, a 2-week mission is planned for 2017. In this mission, scientific studies will be conducted to assess facility capabilities in preparation for longer duration missions. A second follow-on 2-week mission may be planned for early in 2018. In future years, long duration missions of 4, 8 and 12 months are being considered. Missions will include scenarios that simulate for example, transit to and from asteroids, the moon, or other interplanetary travel. Mission operations will be structured to include stressors such as, high workloads, communication delays, and sleep deprivation. Studies completed at the NEK will support International Space Station expeditions, and future exploration missions. Topics studied will include communication, crew autonomy, cultural diversity, human factors, and medical capabilities.

  11. Occupational Radiation Exposure at Commercial Nuclear Power Reactors and Other Facilities 2010, Prepared for the Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 2012

    Energy Technology Data Exchange (ETDEWEB)

    D. E. Lewis D. A. Hagemeyer Y. U. McCormick

    2012-07-07

    This report summarizes the occupational exposure data that are maintained in the U.S. Nuclear Regulatory Commission’s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 2010 annual reports submitted by five of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Because there are no geologic repositories for high-level waste currently licensed and no NRC-licensed low-level waste disposal facilities currently in operation, only five categories will be considered in this report. The annual reports submitted by these licensees consist of radiation exposure records for each monitored individual. These records are analyzed for trends and presented in this report in terms of collective dose and the distribution of dose among the monitored individuals. Annual reports for 2010 were received from a total of 190 NRC licensees. The summation of reports submitted by the 190 licensees indicated that 192,424 individuals were monitored, 81,961 of whom received a measurable dose. When adjusted for transient workers who worked at more than one licensee during the year, there were actually 142,471 monitored individuals and 62,782 who received a measurable dose. The collective dose incurred by these individuals was 10,617 person-rem, which represents a 12% decrease from the 2009 value. This decrease was primarily due to the decrease in collective dose at commercial nuclear power reactors, as well as a decrease in the collective dose for most of the other categories of NRC licensees. The number of individuals receiving a measurable dose also decreased, resulting in an average measurable dose of 0.13 rem for 2010. The average measurable dose is defined as the total effective dose equivalent (TEDE) divided by the number of individuals receiving a measurable dose. In calendar year 2010, the average annual collective dose per reactor for light water reactor

  12. The role of clusters in gas-solids reactors. An experimental study.

    NARCIS (Netherlands)

    Venderbosch, R.H.

    1998-01-01

    This PhD-work is meant to determine the contact efficiency experimentally for fluidization of fine particles over a wide range of superficial gas velocities (dp<200 mm and 0.1

  13. 14MeV neutron irradiation experiment on window materials for fusion experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Fuminobu; Oyama, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iida, Toshiyuki

    1997-06-01

    Data on wavelength spectra of photons emitted from window material during neutron and gamma-ray irradiation has been required for design of next D-T burning fusion reactor such as ITER. Thus, a photon measurement system has been developed to analyze wavelength spectra of photons emitted from the optical window materials during 14MeV-neutron irradiation, and the system consisted of a sample holder, a radiation-resistant optical fiber, a photon counting analyzer and other electronic devices. The irradiation experiments for synthesized sapphire, high-purity silica glass and synthesized quartz were performed using a fusion neutron source FNS. As for all the sample, number of photon emission was proportional to the 14MeV-neutron flux in the range of 10{sup 6}-10{sup 11}n/cm{sup 2}/sec. The photon emission efficiency of F-center luminescence of the sapphire was 2200 {+-} 700photons/MeV, while the efficiency of F{sup +}-center luminescence was two order less than that of F-center. The wavelength spectra of the high-purity silica glass had a large peak around 450nm, which was concerned with decay of self-trapped excitons in oxygen vacancies. Its photon emission efficiency for 14MeV-neutrons has been found to be about 5 {+-} 3photons/MeV in visible range, while that for gamma-rays to be about 135 {+-} 50photons/MeV. The spectrum of photons emitted from the quartz had two large peaks around not only 450nm but also 650nm, and the photon emission efficiency in the wavelength range of 350-750nm was 14 {+-} 4photons/MeV. (author)

  14. The thermal decomposition of the benzyl radical in a heated micro-reactor. I. Experimental findings

    Energy Technology Data Exchange (ETDEWEB)

    Buckingham, Grant T.; Ormond, Thomas K. [Department of Chemistry and Biochemistry, University of Colorado, Boulder, Colorado 80309-0215 (United States); National Renewable Energy Laboratory, 15013 Denver West Parkway, Golden, Colorado 80401 (United States); Porterfield, Jessica P.; Ellison, G. Barney [Department of Chemistry and Biochemistry, University of Colorado, Boulder, Colorado 80309-0215 (United States); Hemberger, Patrick [Molecular Dynamics Group, Paul Scherrer Institut, CH-5232 Villigen-PSI (Switzerland); Kostko, Oleg; Ahmed, Musahid [Chemical Sciences Division, Lawrence Berkeley National Laboratories, Berkeley, California 94720 (United States); Robichaud, David J.; Nimlos, Mark R. [National Renewable Energy Laboratory, 15013 Denver West Parkway, Golden, Colorado 80401 (United States); Daily, John W. [Department of Mechanical Engineering, Center for Combustion and Environmental Research,University of Colorado, Boulder, Colorado 80309-0427 (United States)

    2015-01-28

    The pyrolysis of the benzyl radical has been studied in a set of heated micro-reactors. A combination of photoionization mass spectrometry (PIMS) and matrix isolation infrared (IR) spectroscopy has been used to identify the decomposition products. Both benzyl bromide and ethyl benzene have been used as precursors of the parent species, C{sub 6}H{sub 5}CH{sub 2}, as well as a set of isotopically labeled radicals: C{sub 6}H{sub 5}CD{sub 2}, C{sub 6}D{sub 5}CH{sub 2}, and C{sub 6}H{sub 5}{sup 13}CH{sub 2}. The combination of PIMS and IR spectroscopy has been used to identify the earliest pyrolysis products from benzyl radical as: C{sub 5}H{sub 4}=C=CH{sub 2}, H atom, C{sub 5}H{sub 4}—C ≡ CH, C{sub 5}H{sub 5}, HCCCH{sub 2}, and HC ≡ CH. Pyrolysis of the C{sub 6}H{sub 5}CD{sub 2}, C{sub 6}D{sub 5}CH{sub 2}, and C{sub 6}H{sub 5}{sup 13}CH{sub 2} benzyl radicals produces a set of methyl radicals, cyclopentadienyl radicals, and benzynes that are not predicted by a fulvenallene pathway. Explicit PIMS searches for the cycloheptatrienyl radical were unsuccessful, there is no evidence for the isomerization of benzyl and cycloheptatrienyl radicals: C{sub 6}H{sub 5}CH{sub 2}⇋C{sub 7}H{sub 7}. These labeling studies suggest that there must be other thermal decomposition routes for the C{sub 6}H{sub 5}CH{sub 2} radical that differ from the fulvenallene pathway.

  15. Three-phase packed bed reactor with an evaporating solvent—I. Experimental: the hydrogenation of 2,4,6-trinitrotoluene in methanol

    NARCIS (Netherlands)

    van Gelder, K.B.; Damhof, J.K.; Kroijenga, P.J.; Westerterp, K.R.

    1990-01-01

    In this paper we present experimental data on the three-phase hydrogenation of 2,4,6-trinitrotoluene (TNT) to triaminotoluene. The experiments are performed in a cocurrent upflow packed bed reactor. Methanol is used as an evaporating solvent. The influence of the main operating parameters, the

  16. Optimization of the irradiation beam in the BNCT research facility at IEA-R1 reactor; Otimizacao do feixe de irradiacao na instalacao para estudos em BNCT junto ao reator IEA-R1

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Vinicius Alexandre de

    2014-07-01

    Boron Neutron Capture Therapy (BNCT) is a radiotherapeutic technique for the treatment of some types of cancer whose useful energy comes from a nuclear reaction that occurs when thermal neutron impinges upon a Boron-10 atom. In Brazil there is a research facility built along the beam hole number 3 of the IEA-R1 research reactor at IPEN, which was designed to perform BNCT research experiments. For a good performance of the technique, the irradiation beam should be mostly composed of thermal neutrons with a minimum as possible gamma and above thermal neutron components. This work aims to monitor and evaluate the irradiation beam on the sample irradiation position through the use of activation detectors (activation foils) and also to propose, through simulation using the radiation transport code, MCNP, new sets of moderators and filters which shall deliver better irradiation fields at the irradiation sample position In this work, a simulation methodology, based on a MCNP card, known as wwg (weight window generation) was studied, and the neutron energy spectrum has been experimentally discriminated at 5 energy ranges by using a new set o activation foils. It also has been concluded that the BNCT research facility has the required thermal neutron flux to perform studies in the area and it has a great potential for improvement for tailoring the irradiation field. (author)

  17. An Experimental Investigation of Sewage Sludge Gasification in a Fluidized Bed Reactor

    Science.gov (United States)

    Calvo, L. F.; García, A. I.; Otero, M.

    2013-01-01

    The gasification of sewage sludge was carried out in a simple atmospheric fluidized bed gasifier. Flow and fuel feed rate were adjusted for experimentally obtaining an air mass : fuel mass ratio (A/F) of 0.2 gasification. This allowed improving the process heat transfer and, therefore, gasification efficiency. The heating value of the produced gas was 8.4 MJ/Nm, attaining a hot gas efficiency of 70% and a cold gas efficiency of 57%. PMID:24453863

  18. The NASA Ames Hypervelocity Free Flight Aerodynamic Facility: Experimental Simulation of the Atmospheric Break-Up of Meteors

    Science.gov (United States)

    Wilder, M. C.; Bogdanoff, D. W.

    2015-01-01

    The Hypervelocity Free Flight Aerodynamic Facility at NASA Ames Research Center provides a potential platform for the experimental simulation of meteor breakup at conditions that closely match full-scale entry condition for select parameters. The poster describes the entry environment simulation capabilities of the Hypervelocity Free Flight Aerodynamic Facility (HFFAF) at NASA Ames Research Center and provides example images of the fragmentation of a hypersonic projectile for which break-up was initiated by mechanical forces (impact with a thin polymer diaphragm).

  19. Velo and REXAN - Integrated Data Management and High Speed Analysis for Experimental Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Kleese van Dam, Kerstin; Carson, James P.; Corrigan, Abigail L.; Einstein, Daniel R.; Guillen, Zoe C.; Heath, Brandi S.; Kuprat, Andrew P.; Lanekoff, Ingela T.; Lansing, Carina S.; Laskin, Julia; Li, Dongsheng; Liu, Yan; Marshall, Matthew J.; Miller, Erin A.; Orr, Galya; Pinheiro da Silva, Paulo; Ryu, Seun; Szymanski, Craig J.; Thomas, Mathew

    2013-01-10

    The Chemical Imaging Initiative at the Pacific Northwest National Laboratory (PNNL) is creating a ‘Rapid Experimental Analysis’ (REXAN) Framework, based on the concept of reusable component libraries. REXAN allows developers to quickly compose and customize high throughput analysis pipelines for a range of experiments, as well as supporting the creation of multi-modal analysis pipelines. In addition, PNNL has coupled REXAN with its collaborative data management and analysis environment Velo to create an easy to use data management and analysis environments for experimental facilities. This paper will discuss the benefits of Velo and REXAN in the context of three examples: PNNL High Resolution Mass Spectrometry - reducing analysis times from hours to seconds, and enabling the analysis of much larger data samples (100KB to 40GB) at the same time · ALS X-Ray tomography - reducing analysis times of combined STXM and EM data collected at the ALS from weeks to minutes, decreasing manual work and increasing data volumes that can be analysed in a single step ·Multi-modal nano-scale analysis of STXM and TEM data - providing a semi automated process for particle detection The creation of REXAN has significantly shortened the development time for these analysis pipelines. The integration of Velo and REXAN has significantly increased the scientific productivity of the instruments and their users by creating easy to use data management and analysis environments with greatly reduced analysis times and improved analysis capabilities.

  20. System integration of RF based negative ion experimental facility at IPR

    Energy Technology Data Exchange (ETDEWEB)

    Bansal, G; Bandyopadhyay, M; Singh, M J; Gahlaut, A; Soni, J; Pandya, K; Parmar, K G; Sonara, J; Chakraborty, A, E-mail: bansal@ipr.res.i [Institute for Plasma Research, Bhat, Gandhinagar (Gujarat) 382 428 (India)

    2010-02-01

    The setting up of RF based negative ion experimental facility shall witness the beginning of experiments on the negative ion source fusion applications in India. A 1 MHz RF generator shall launch 100 kW RF power into a single driver on the plasma source to produce a plasma of density {approx}5 x 10{sup 12} cm{sup -3}. The source can deliver a negative ion beam of {approx}10 A with a current density of {approx}30 mA/cm{sup 2} and accelerated to 35 kV through an electrostatic ion accelerator. The experimental system is similar to a RF based negative ion source, BATMAN, presently operating at IPP. The subsystems for source operation are designed and procured principally from indigenous resources, keeping the IPP configuration as a base line. The operation of negative ion source is supported by many subsystems e.g. vacuum pumping system with gate valves, cooling water system, gas feed system, cesium delivery system, RF generator, high voltage power supplies, data acquisition and control system, and different diagnostics. The first experiments of negative ion source are expected to start at IPR from the middle of 2009.

  1. Proton beam characterization in the experimental room of the Trento Proton Therapy facility

    Science.gov (United States)

    Tommasino, F.; Rovituso, M.; Fabiano, S.; Piffer, S.; Manea, C.; Lorentini, S.; Lanzone, S.; Wang, Z.; Pasini, M.; Burger, W. J.; La Tessa, C.; Scifoni, E.; Schwarz, M.; Durante, M.

    2017-10-01

    As proton therapy is becoming an established treatment methodology for cancer patients, the number of proton centres is gradually growing worldwide. The economical effort for building these facilities is motivated by the clinical aspects, but might be also supported by the potential relevance for the research community. Experiments with high-energy protons are needed not only for medical physics applications, but represent also an essential part of activities dedicated to detector development, space research, radiation hardness tests, as well as of fundamental research in nuclear and particle physics. Here we present the characterization of the beam line installed in the experimental room of the Trento Proton Therapy Centre (Italy). Measurements of beam spot size and envelope, range verification and proton flux were performed in the energy range between 70 and 228 MeV. Methods for reducing the proton flux from typical treatments values of 106-109 particles/s down to 101-105 particles/s were also investigated. These data confirm that a proton beam produced in a clinical centre build by a commercial company can be exploited for a broad spectrum of experimental activities. The results presented here will be used as a reference for future experiments.

  2. Prolongation of the BOR-60 reactor operation

    Directory of Open Access Journals (Sweden)

    Alexey l. Izhutov

    2015-04-01

    Full Text Available The fast neutron reactor BOR-60 is one of the key experimental facilities worldwide to perform large-scale tests of fuel, absorbing, and structural materials for advanced reactors. The BOR-60 reactor was put into operation in December 1969, and by the end of 2014 it had been operating on power for ∼265,000 hours. BOR-60 still demonstrates potential capabilities to extend the lifetime of sodium-cooled fast reactors. The BOR-60 lifetime should have expired at the end of 2014. Over the past few years, a great scope of work has been performed to justify the possibility of extending its lifetime. The work included inspection of the equipment conditions, calculations and experimental research on operating parameters and the conditions of nonremovable components, investigation of the structural material samples after their long-term operation under irradiation, etc. Based on the results of the work performed, the residual lifetime was evaluated and the reactor operator made a decision to extend the lifetime period of the BOR-60 reactor. After considering both a set of documents about the reactor conditions and the positive decision of independent experts, the Regulatory Authority of the Russian Federation extended the BOR-60 operating license up to 2020.

  3. Local Neutron Flux Distribution Measurements by Wire-Dosimetry in the AMMON Experimental Program in the EOLE Reactor

    Directory of Open Access Journals (Sweden)

    Gruel A.

    2016-01-01

    Full Text Available Dosimetry measurements were carried out during the AMMON experimental program, in the EOLE facility. Al-0.1 wt% Au wires were positioned along curved fuel plates of JHR-type assemblies to investigate the azimuthal and axial gold capture rate profiles, directly linked to the thermal and epithermal flux. After irradiation, wires were cut into small segments (a few mm, and the gold capture rate of each part was measured by gamma spectrometry on the MADERE platform. This paper presents results in the “hafnium” configuration, and more specifically the azimuthal flux profile characterization. The final uncertainty on each measured wire lies below 1% (at 2 standard deviations. Experimental profiles are in a good agreement against Monte Carlo calculations, and the 4% capture rate increase at the plate edge is well observed. The flux dissymmetry due to assembly position in the core is also measured, and shows a 10% discrepancy between the two edges of the plate.

  4. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    Energy Technology Data Exchange (ETDEWEB)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  5. An analytical and experimental investigation of natural circulation transients in a model pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Massoud, M

    1987-01-01

    Natural Circulation phenomena in a simulated PWR was investigated experimentally and analytically. The experimental investigation included determination of system characteristics as well as system response to the imposed transient under symmetric and asymmetric operations. System characteristics were used to obtain correlation for heat transfer coefficient in heat exchangers, system flow resistance, and system buoyancy heat. Asymmetric transients were imposed to study flow oscillation and possible instability. The analytical investigation encompassed development of mathematical model for single-phase, steady-state and transient natural circulation as well as modification of existing model for two-phase flow analysis of phenomena such as small break LOCA, high pressure coolant injection and pump coast down. The developed mathematical model for single-phase analysis was computer coded to simulate the imposed transients. The computer program, entitled ''Symmetric and Asymmetric Analysis of Single-Phase Flow (SAS),'' were employed to simulate the imposed transients. It closely emulated the system behavior throughout the transient and subsequent steady-state. Modifications for two-phase flow analysis included addition of models for once-through steam generator and electric heater rods. Both programs are faster than real time. Off-line, they can be used for prediction and training applications while on-line they serve for simulation and signal validation. The programs can also be used to determine the sensitivity of natural circulation behavior to variation of inputs such as secondary distribution and power transients.

  6. Resistance to experimental autoimmune encephalomyelitis development in Lewis rats from a conventional animal facility

    Directory of Open Access Journals (Sweden)

    Sofia Fernanda Gonçalves Zorzella

    2007-12-01

    Full Text Available Experimental autoimmune encephalomyelitis (EAE is an inflammatory disease of the brain and spinal cord that is mediated by CD4+ T lymphocytes specific to myelin components. In this study we compared development of EAE in Lewis rats from two colonies, one kept in pathogen-free conditions (CEMIB colony and the other (Botucatu colony kept in a conventional animal facility. Female Lewis rats were immunized with 100 µl of an emulsion containing 50 µg of myelin, associated with incomplete Freund's adjuvant plus Mycobacterium butyricum. Animals were daily evaluated for clinical score and weight. CEMIB colony presented high EAE incidence with clinical scores that varied from three to four along with significant weight losses. A variable disease incidence was observed in the Botucatu colony with clinical scores not higher than one and no weight loss. Immunological and histopathological characteristics were also compared after 20 days of immunization. Significant amounts of IFN-gamma, TNF-alpha and IL-10 were induced by myelin in cultures from CEMIB animals but not from the Botucatu colony. Significantly higher levels of anti-myelin IgG1 were detected in the CEMIB colony. Clear histopathological differences were also found. Cervical spinal cord sections from CEMIB animals showed typical perivascular inflammatory foci whereas samples from the Botucatu colony showed a scanty inflammatory infiltration. Helminths were found in animals from Botucatu colony but not, as expected, in the CEMIB pathogen-free animals. As the animals maintained in a conventional animal facility developed a very discrete clinical, and histopathological EAE in comparison to the rats kept in pathogen-free conditions, we believe that environmental factors such as intestinal parasites could underlie this resistance to EAE development, supporting the applicability of the hygiene hypothesis to EAE.

  7. Radiation Damage in Nuclear Fuel for Advanced Burner Reactors: Modeling and Experimental Validation

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, Niels Gronbech; Asta, Mark; Ozolins, Nigel Browning' Vidvuds; de Walle, Axel van; Wolverton, Christopher

    2011-12-29

    The consortium has completed its existence and we are here highlighting work and accomplishments. As outlined in the proposal, the objective of the work was to advance the theoretical understanding of advanced nuclear fuel materials (oxides) toward a comprehensive modeling strategy that incorporates the different relevant scales involved in radiation damage in oxide fuels. Approaching this we set out to investigate and develop a set of directions: 1) Fission fragment and ion trajectory studies through advanced molecular dynamics methods that allow for statistical multi-scale simulations. This work also includes an investigation of appropriate interatomic force fields useful for the energetic multi-scale phenomena of high energy collisions; 2) Studies of defect and gas bubble formation through electronic structure and Monte Carlo simulations; and 3) an experimental component for the characterization of materials such that comparisons can be obtained between theory and experiment.

  8. Reactor Safety Research Programs Quarterly Report April- June 1981

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1981-09-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL} from April1 through June 30, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory {INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  9. Chemical and Radiochemical Constituents in Water from Wells in the Vicinity of the Naval Reactors Facility, Idaho National Engineering and Environmental Laboratory, Idaho, 1996

    Energy Technology Data Exchange (ETDEWEB)

    L. L. Knobel; R. C. Bartholomay; B. J. Tucker; L. M. Williams (USGS)

    1999-10-01

    The U.S. Geological Survey, in response to a request from the U.S. Department of Energy's Pittsburgh Naval Reactors Office, Idaho Branch Office (IBO), samples water from 13 wells during 1996 as part of a long-term project to monitor water quality to the Snake River Plain aquifer in the vicinity of the Naval Reactors Facility (NRF), Idaho National Engineering and Environmental Laboratory, Idaho. The IBO requires information about the mobility of radionuclide- and chemical-waste constituents in the Snake River Plain aquifer. Waste-constituent mobility is determined principally by (1) the rate and direction of ground-water flow; (2) the locations, quantities, and methods of waste disposal; (3) waste-constituents chemistry; and (4) the geochemical processes taking place in the aquifer. The purpose of the data-collection program is to provide IBO with water-chemistry data to evaluate the effect of NRF activities on the water quality of the Snake River Plain aquifer. Water samples were analyzed for naturally occurring constituents and man-made contaminants.

  10. Chemical constituents in water from wells in the vicinity of the Naval Reactors Facility, Idaho National Engineering Laboratory, Idaho, 1991--93

    Energy Technology Data Exchange (ETDEWEB)

    Tucker, B.J.; Knobel, L.L.; Bartholomay, R.C.

    1995-11-01

    The US Geological Survey, in response to a request from the US Department of Energy`s Pittsburgh Naval Reactors Office, Idaho Branch Office, sampled 14 wells during 1991--93 as part of a long-term project to monitor water quality of the Snake River Plain aquifer in the vicinity of the Naval Reactors Facility, Idaho National Engineering Laboratory, Idaho. Water samples were analyzed for manmade contaminants and naturally occurring constituents. One hundred sixty-one samples were collected from 10 ground-water monitoring wells and 4 production wells. Twenty-one quality-assurance samples also were collected and analyzed; 2 were blank samples and 19 were replicate samples. The two blank samples contained concentrations of six inorganic constituents that were slightly greater than the laboratory reporting levels (the smallest measured concentration of a constituent that can be reported using a given analytical method). Concentrations of other constituents in the blank samples were less than their respective reporting levels. The 19 replicate samples and their respective primary samples generated 614 pairs of analytical results for a variety of chemical and radiochemical constituents. Of the 614 data pairs, 588 were statistically equivalent at the 95% confidence level; about 96% of the analytical results were in agreement. Two pairs of turbidity measurements were not evaluated because of insufficient information and one primary sample collected in January 1992 contained tentatively identified organic compounds when the replicate sample did not.

  11. Experimental measurement and Monte Carlo assessment of Argon-41 production in a PET cyclotron facility.

    Science.gov (United States)

    Infantino, Angelo; Valtieri, Lorenzo; Cicoria, Gianfranco; Pancaldi, Davide; Mostacci, Domiziano; Marengo, Mario

    2015-12-01

    In a medical cyclotron facility, (41)Ar (t1/2 = 109.34 m) is produced by the activation of air due to the neutron flux during irradiation, according to the (40)Ar(n,γ)(41)Ar reaction; this is particularly relevant in widely diffused high beam current cyclotrons for the production of PET radionuclides. While theoretical estimations of the (41)Ar production have been published, no data are available on direct experimental measurements for a biomedical cyclotron. In this work, we describe a sampling methodology and report the results of an extensive measurement campaign. Furthermore, the experimental results are compared with Monte Carlo simulations performed with the FLUKA code. To measure (41)Ar activity, air samples were taken inside the cyclotron bunker in sealed Marinelli beakers, during the routine production of (18)F with a 16.5 MeV GE-PETtrace cyclotron; this sampling thus reproduces a situation of absence of air changes. Samples analysis was performed in a gamma-ray spectrometry system equipped with HPGe detector. Monte Carlo assessment of the (41)Ar saturation yield was performed directly using the standard FLUKA score RESNUCLE, and off-line by the convolution of neutron fluence with cross section data. The average (41)Ar saturation yield per one liter of air of (41)Ar, measured in gamma-ray spectrometry, resulted to be 3.0 ± 0.6 Bq/µA*dm(3) while simulations gave a result of 6.9 ± 0.3 Bq/µA*dm(3) in the direct assessment and 6.92 ± 0.22 Bq/µA*dm(3) by the convolution neutron fluence-to-cross section. Copyright © 2015 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.

  12. Flowing lead spallation target design for use in an ADTT experimental facility located at LAMPF

    Science.gov (United States)

    Beard, C. A.; Bracht, R. R.; Buksa, J. J.; Chaves, W.; DeVolder, B. G.; O'Brien, H.; Park, J. J.; Parker, R. B.; Pillai, C.; Potter, R. C.; Reid, R. S.; Trujillo, D. A.; Vela, O. A.; Venneri, F.; Weinacht, D. J.; Wender, S. A.; Wilson, W. B.; Woloshun, K. A.

    1995-09-01

    A conceptual design has been initiated for a flowing lead spallation target for use in an ADTT experimental facility located at LAMPF. The lead is contained using Nb-1Zr as the structural material. This material was selected based on its favorable material properties as well as its compatibility with the flowing lead. Heat deposited in the lead and the Nb-1Zr container by the 800-MeV, 1-mA beam is removed by the flowing lead and transferred to helium via a conventional heat exchanger. The neutronic, thermal hydraulic, and stress characteristics of the system have been determined. In addition, a module to control the thaw and freeze of the lead has been developed and incorporated into the target system design. The entire primary target system (spallation target, thaw/freeze system, and intermediate heat exchanger) has been designed to be built as a contained module to allow easy insertion into an experimental ADTT blanket assembly and to provide multiple levels of containment for the lead. For the 800-MeV LAMPF beam, the target delivers a source of approximately 18 neutrons/proton. A total of 540 kW are deposited in the target. The lead temperature ranges from 400 to 500 C. The peak structural heating occurs at the beam interface, and the target is designed to maximize cooling at this point. An innovative thin-window structure has been incorporated that allows direct, convective cooling of the window by the inlet flowing lead. Safe and reliable operation of the target has been maximized through simple, robust engineering.

  13. Contribution to modeling of the reflooding of a severely damaged reactor core using PRELUDE experimental results

    Energy Technology Data Exchange (ETDEWEB)

    Bachrata, A.; Fichot, F.; Repetto, G. [Institut de Radioprotection et de Surete Nucleaire IRSN, Cadarache (France); Quintard, M. [Universite de Toulouse, INPT, UPS, IMFT Institut de Mecanique des Fluides de Toulouse, Allee Camille Soula, F-31400 Toulouse (France); CNRS, IMFT, F-31400 Toulouse (France); Fleurot, J. [Institut de Radioprotection et de Surete Nucleaire IRSN, Cadarache (France)

    2012-07-01

    In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. The reflooding (injection of water into core) may be applied if the availability of safety injection is recovered during accident. If the injection becomes available only in the late phase of accident, water will enter a core configuration that will differ significantly from original rod-bundle geometry. Any attempt to inject water after significant core degradation can lead to further fragmentation of core material. The fragmentation of fuel rods may result in the formation of a 'debris bed'. The typical particle size in a debris bed might reach few millimeters (characteristic length-scale: 1 to 5 mm), i.e., a high permeability porous medium. The French 'Institut de Radioprotection et de Surete Nucleaire' is developing experimental programs (PEARL and PRELUDE) and simulation tools (ICARE-CATHARE and ASTEC) to study and optimize the severe accident management strategy and to assess the probabilities to stop the progress of in-vessel core degradation. It is shown that the quench front exhibits either a ID behaviour or a 2D one, depending on injection rate or bed characteristics. The PRELUDE experiment covers a rather large range of variation of parameters, for which the developed model appears to be quite predictive. (authors)

  14. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    Science.gov (United States)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  15. High power 1 MeV neutral beam system and its application plan for the international tokamak experimental reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hemsworth, R.S. [ITER Joint Central Team, Naka, Ibaraki (Japan)

    1997-03-01

    This paper describes the Neutral Beam Injection system which is presently being designed for the International Tokamak Experimental Reactor, ITER, in Europe Japan and Russia, with co-ordination by the Joint Central Team of ITER at Naka, Japan. The proposed system consists of three negative ion based neutral injectors, delivering a total of 50 MW of 1 MeV D{sup 0} to the ITER plasma for a pulse length of >1000 s. Each injectors uses a single caesiated volume arc discharge negative ion source, and a multi-grid, multi-aperture accelerator, to produce about 40 A of 1 MeV D{sup -}. This will be neutralized by collisions with D{sub 2} in a sub-divided gas neutralizer, which has a conversion efficiency of about 60%. The charged fraction of the beam emerging from the neutralizer is dumped in an electrostatic residual ion dump. A water cooled calorimeter can be moved into the beam path to intercept the neutral beam, allowing commissioning of the injector independent of ITER. ITER is scheduled to produce its first plasma at the beginning of 2008, and the planning of the R and D, construction and installation foresees the neutral injection system being available from the start of ITER operations. (author)

  16. Modeling of secondary emission processes in the negative ion based electrostatic accelerator of the International Thermonuclear Experimental Reactor

    Directory of Open Access Journals (Sweden)

    G. Fubiani

    2008-01-01

    Full Text Available The negative ion electrostatic accelerator for the neutral beam injector of the International Thermonuclear Experimental Reactor (ITER is designed to deliver a negative deuterium current of 40 A at 1 MeV. Inside the accelerator there are several types of interactions that may create secondary particles. The dominating process originates from the single and double stripping of the accelerated negative ion by collision with the residual molecular deuterium gas (≃29% losses. The resulting secondary particles (positive ions, neutrals, and electrons are accelerated and deflected by the electric and magnetic fields inside the accelerator and may induce more secondaries after a likely impact with the accelerator grids. This chain of reactions is responsible for a non-negligible heat load on the grids and must be understood in detail. In this paper, we will provide a comprehensive summary of the physics involved in the process of secondary emission in a typical ITER-like negative ion electrostatic accelerator together with a precise description of the numerical method and approximations involved. As an example, the multiaperture-multigrid accelerator concept will be discussed.

  17. Analysis of quench-vent pressures for present design of ITER (International Thermonuclear Experimental Reactor) TF (toroidal field) coils

    Energy Technology Data Exchange (ETDEWEB)

    Slack, D.S.

    1989-09-20

    The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Japan, the European Community, the Union of the Soviet Union, and the United States. This paper examines the effects of a quench within the toroidal field (TF) coils based on current ITER design. It is a preliminary, rough analysis. Its intent is to assist ITER designers while more accurate computer codes are being developed and to provide a check against these more rigorous solutions. Rigorous solutions to the quench problem are very complex involving three- dimensional heat transfer, extreme changes in heat capacities and copper resistivity, and varying flow dynamics within the conductors. This analysis addresses all these factors in an approximate way. The result is much less accurate than a rigorous analysis. Results here could be in error as much as 30 to 40 percent. However, it is believed that this paper can still be very useful to the coil designer. Coil pressures and temperatures vs time into a quench are presented. Rate of helium vent, energy deposition in the coil, and depletion of magnetic stored energy are also presented. Peak pressures are high (about 43 MPa). This is due to the very long vent path length (446 m), small hydraulic diameters, and high current densities associated with ITER's cable-in-conduit design. The effects of these pressures as well as the ability of the coil to be self protecting during a quench are discussed. 3 refs., 1 fig., 3 tabs.

  18. Recent progress on the hydrogen storage properties of ZrCo-based alloys applied in International Thermonuclear Experimental Reactor (ITER

    Directory of Open Access Journals (Sweden)

    Feng Wang

    2017-02-01

    Full Text Available Under the development of International Thermonuclear Experimental Reactor (ITER system aimed at realizing the controllable fusion reaction to solve the energy crisis fundamentally, there is an urgent need to find an appropriate material for tritium handling. ZrCo alloy is considered to be a promising candidate for the storage and delivery of hydrogen isotopes due to the favorable characteristics such as low plateau pressure for absorption, high dissociation pressure at moderate temperature and better ability of trapping 3He. However, the hydrogen induced disproportionation and the slower recovery/deliverty rate of ZrCo-based alloys have limited their further application in ITER system. This review summarizes the efforts towards enhancing the hydrogen storage properties of ZrCo-based alloys including element substitution, surface modification, disproportionation mechanism investigation and the isotope effect study. Element substitution and surface modification play positive role to improve the anti-disproportionation ability and kinetic property of the alloys. However, the ZrCo-based alloys require to be further modified by more attempts such as new composition, novelty modification method or catalyst addition in order to better satisfy the application demands for tritium handling. Moreover, new insight for further understanding the inner disproportionation mechanisms of this material is needed by combining the advance characterization and theoretical analysis, which is in favor of addressing the disproportionation problem of the ZrCo-based alloys essentially.

  19. Experimental engineering section off-gas decontamination facility's fractionator column: installation and performance

    Energy Technology Data Exchange (ETDEWEB)

    Gilliam, T. M.; Fowler, V. L.; Inman, D. J.

    1978-03-01

    A detailed description of the third column recently installed in the Experimental Engineering Section Off-Gas Decontamination Facility (EES-ODF) is presented. The EES-ODF is being used to provide engineering-scale experiments (nominal gas and liquid flows of 5 scfm and 0.5 gpm, respectively) in the development of the Krypton Absorption in Liquid CO/sub 2/ (KALC) process. A detailed discussion of the column's construction is provided. This discussion includes the peripherals associated with the column, such as refrigeration, heat exchangers, instrumentation, etc. The compressibility of Goodloe packing (the packing in the other columns) and the possible reduced throughput due to this compression have revealed the desirablility of a random (i.e., noncompressible) packing. Toward this end, the third column is packed with a new random packing (PRO-PAK). A preliminary comparison between this packing and the woven wire mesh packing (Goodloe) used in the other two columns has been made. Experiments comparing the throughput capacity indicate that the PRO-PAK packing has approximately 60% the capacity of Goodloe for a CO/sub 2/ system. When used as a fractionator or stripper with the basic O/sub 2/-Kr-CO/sub 2/ KALC system, the PRO-PAK column produced HTU values less than or equal to the GOODLOE columns under similar operating conditions.

  20. A Proof-of-Concept for Semantically Interoperable Federation of IoT Experimentation Facilities

    Directory of Open Access Journals (Sweden)

    Jorge Lanza

    2016-06-01

    Full Text Available The Internet-of-Things (IoT is unanimously identified as one of the main pillars of future smart scenarios. The potential of IoT technologies and deployments has been already demonstrated in a number of different application areas, including transport, energy, safety and healthcare. However, despite the growing number of IoT deployments, the majority of IoT applications tend to be self-contained, thereby forming application silos. A lightweight data centric integration and combination of these silos presents several challenges that still need to be addressed. Indeed, the ability to combine and synthesize data streams and services from diverse IoT platforms and testbeds, holds the promise to increase the potentiality of smart applications in terms of size, scope and targeted business context. In this article, a proof-of-concept implementation that federates two different IoT experimentation facilities by means of semantic-based technologies will be described. The specification and design of the implemented system and information models will be described together with the practical details of the developments carried out and its integration with the existing IoT platforms supporting the aforementioned testbeds. Overall, the system described in this paper demonstrates that it is possible to open new horizons in the development of IoT applications and experiments at a global scale, that transcend the (silo boundaries of individual deployments, based on the semantic interconnection and interoperability of diverse IoT platforms and testbeds.

  1. EXPERIMENTAL AND COMPUTATIONAL ACTIVITIES AT THE OREGON STATE UNIVERSITY NEES TSUNAMI RESEARCH FACILITY

    Directory of Open Access Journals (Sweden)

    S.C. Yim

    2009-01-01

    Full Text Available A diverse series of research projects have taken place or are underway at the NEES Tsunami Research Facility at Oregon State University. Projects range from the simulation of the processes and effects of tsunamis generated by sub-aerial and submarine landslides (NEESR, Georgia Tech., model comparisons of tsunami wave effects on bottom profiles and scouring (NEESR, Princeton University, model comparisons of wave induced motions on rigid and free bodies (Shared-Use, Cornell, numerical model simulations and testing of breaking waves and inundation over topography (NEESR, TAMU, structural testing and development of standards for tsunami engineering and design (NEESR, University of Hawaii, and wave loads on coastal bridge structures (non-NEES, to upgrading the two-dimensional wave generator of the Large Wave Flume. A NEESR payload project (Colorado State University was undertaken that seeks to improve the understanding of the stresses from wave loading and run-up on residential structures. Advanced computational tools for coupling fluid-structure interaction including turbulence, contact and impact are being developed to assist with the design of experiments and complement parametric studies. These projects will contribute towards understanding the physical processes that occur during earthquake generated tsunamis including structural stress, debris flow and scour, inundation and overland flow, and landslide generated tsunamis. Analytical and numerical model development and comparisons with the experimental results give engineers additional predictive tools to assist in the development of robust structures as well as identification of hazard zones and formulation of hazard plans.

  2. High-Resolution Melting Curve Analysis for Identification of Pasteurellaceae Species in Experimental Animal Facilities.

    Science.gov (United States)

    Miller, Manuel; Zorn, Julia; Brielmeier, Markus

    2015-01-01

    Pasteurellaceae are among the most prevalent bacterial pathogens isolated from mice housed in experimental animal facilities. Reliable detection and differentiation of Pasteurellaceae are essential for high-quality health monitoring. In this study, we combined a real-time PCR assay amplifying a variable region in the 16S rRNA sequence with high-resolution melting curve analysis (HRM) to identify and differentiate among the commonly isolated species Pasteurella pneumotropica biotypes "Jawetz" and "Heyl", Actinobacillus muris, and Haemophilus influenzaemurium. We used a set of six reference strains for assay development, with the melting profiles of these strains clearly distinguishable due to DNA sequence variations in the amplicon. For evaluation, we used real-time PCR/HRM to test 25 unknown Pasteurellaceae isolates obtained from an external diagnostic laboratory and found the results to be consistent with those of partial 16S rRNA sequencing. The real-time PCR/HRM method provides a sensitive, rapid, and closed-tube approach for Pasteurellaceae species identification for health monitoring of laboratory mice.

  3. High-Resolution Melting Curve Analysis for Identification of Pasteurellaceae Species in Experimental Animal Facilities.

    Directory of Open Access Journals (Sweden)

    Manuel Miller

    Full Text Available Pasteurellaceae are among the most prevalent bacterial pathogens isolated from mice housed in experimental animal facilities. Reliable detection and differentiation of Pasteurellaceae are essential for high-quality health monitoring. In this study, we combined a real-time PCR assay amplifying a variable region in the 16S rRNA sequence with high-resolution melting curve analysis (HRM to identify and differentiate among the commonly isolated species Pasteurella pneumotropica biotypes "Jawetz" and "Heyl", Actinobacillus muris, and Haemophilus influenzaemurium. We used a set of six reference strains for assay development, with the melting profiles of these strains clearly distinguishable due to DNA sequence variations in the amplicon. For evaluation, we used real-time PCR/HRM to test 25 unknown Pasteurellaceae isolates obtained from an external diagnostic laboratory and found the results to be consistent with those of partial 16S rRNA sequencing. The real-time PCR/HRM method provides a sensitive, rapid, and closed-tube approach for Pasteurellaceae species identification for health monitoring of laboratory mice.

  4. Experimental neutron capture data of $^{58}$Ni from the CERN n_TOF facility

    CERN Document Server

    Žugec, P.; Colonna, N.; Bosnar, D.; Altstadt, S.; Andrzejewski, J.; Audouin, L.; Bécares, V.; Bečvář, F.; Belloni, F.; Berthoumieux, E.; Billowes, J.; Boccone, V.; Brugger, M.; Calviani, M.; Calviño, F.; Cano-Ott, D.; Carrapiço, C.; Cerutti, F.; Chiaveri, E.; Chin, M.; Cortés, G.; Cortés-Giraldo, M.A.; Diakaki, M.; Domingo-Pardo, C.; Duran, I.; Dzysiuk, N.; Eleftheriadis, C.; Ferrari, A.; Fraval, K.; Ganesan, S.; García, A.R.; Giubrone, G.; Gómez-Hornillos, M.B.; Gonçalves, I.F.; González-Romero, E.; Griesmayer, E.; Guerrero, C.; Gunsing, F.; Gurusamy, P.; Jenkins, D.G.; Jericha, E.; Kadi, Y.; Käppeler, F.; Karadimos, D.; Koehler, P.; Kokkoris, M.; Krtička, M.; Kroll, J.; Langer, C.; Lederer, C.; Leeb, H.; Leong, L.S.; Losito, R.; Manousos, A.; Marganiec, J.; Martìnez, T.; Massimi, C.; Mastinu, P.F.; Mastromarco, M.; Meaze, M.; Mendoza, E.; Mengoni, A.; Milazzo, P.M.; Mingrone, F.; Mirea, M.; Mondalaers, W.; Paradela, C.; Pavlik, A.; Perkowski, J.; Pignatari, M.; Plompen, A.; Praena, J.; Quesada, J.M.; Rauscher, T.; Reifarth, R.; Riego, A.; Roman, F.; Rubbia, C.; Sarmento, R.; Schillebeeckx, P.; Schmidt, S.; Tagliente, G.; Tain, J.L.; Tarrío, D.; Tassan-Got, L.; Tsinganis, A.; Valenta, S.; Vannini, G.; Variale, V.; Vaz, P.; Ventura, A.; Versaci, R.; Vermeulen, M.J.; Vlachoudis, V.; Vlastou, R.; Wallner, A.; Ware, T.; Weigand, M.; Weiß, C.; Wright, T.

    2013-01-01

    The $^{58}$Ni $(n,\\gamma)$ cross section has been measured at the neutron time of flight facility n_TOF at CERN, in the energy range from 27 meV up to 400 keV. In total, 51 resonances have been analyzed up to 122 keV. Maxwellian averaged cross sections (MACS) have been calculated for stellar temperatures of kT$=$5-100 keV with uncertainties of less than 6%, showing fair agreement with recent experimental and evaluated data up to kT = 50 keV. The MACS extracted in the present work at 30 keV is 34.2$\\pm$0.6$_\\mathrm{stat}\\pm$1.8$_\\mathrm{sys}$ mb, in agreement with latest results and evaluations, but 12% lower relative to the recent KADoNIS compilation of astrophysical cross sections. When included in models of the s-process nucleosynthesis in massive stars, this change results in a 60% increase of the abundance of $^{58}$Ni, with a negligible propagation on heavier isotopes. The reason is that, using both the old or the new MACS, 58Ni is efficiently depleted by neutron captures.

  5. On Using Shaped Honeycombs for Experimental Generation of Arbitrary Velocity Profiles in Test Facilities

    Science.gov (United States)

    Safaripour, Alireza; Olson, David; Naguib, Ahmed; Koochesfahani, Manoochehr

    2016-11-01

    It is common to use a uniform approach flow in the study of most problems in aerodynamics. Motivated by situations where the approach flow is not uniform, the focus of the current work is on the experimental generation of arbitrary velocity profiles in a flow facility (water tunnel) using the shaped honeycomb technique originally proposed by Kotansky (1966). Employing further refinement of this approach, multiple honeycomb devices are designed and fabricated to produce prescribed velocity profiles. The performance of these devices is assessed in terms of their agreement with the desired velocity profiles and the level of turbulence they produce. Single-component molecular tagging velocimetry (1c-MTV) is used to characterize the resulting mean and fluctuating streamwise velocity profiles and their streamwise development. The shaped honeycomb technique is shown to be effective in producing the desired velocity profiles with high fidelity while maintaining velocity fluctuations level at or below that of the freestream prior to installation of the devices. This work was supported by AFOSR Award Number FA9550-15-1-0224.

  6. A Proof-of-Concept for Semantically Interoperable Federation of IoT Experimentation Facilities.

    Science.gov (United States)

    Lanza, Jorge; Sanchez, Luis; Gomez, David; Elsaleh, Tarek; Steinke, Ronald; Cirillo, Flavio

    2016-06-29

    The Internet-of-Things (IoT) is unanimously identified as one of the main pillars of future smart scenarios. The potential of IoT technologies and deployments has been already demonstrated in a number of different application areas, including transport, energy, safety and healthcare. However, despite the growing number of IoT deployments, the majority of IoT applications tend to be self-contained, thereby forming application silos. A lightweight data centric integration and combination of these silos presents several challenges that still need to be addressed. Indeed, the ability to combine and synthesize data streams and services from diverse IoT platforms and testbeds, holds the promise to increase the potentiality of smart applications in terms of size, scope and targeted business context. In this article, a proof-of-concept implementation that federates two different IoT experimentation facilities by means of semantic-based technologies will be described. The specification and design of the implemented system and information models will be described together with the practical details of the developments carried out and its integration with the existing IoT platforms supporting the aforementioned testbeds. Overall, the system described in this paper demonstrates that it is possible to open new horizons in the development of IoT applications and experiments at a global scale, that transcend the (silo) boundaries of individual deployments, based on the semantic interconnection and interoperability of diverse IoT platforms and testbeds.

  7. REVIEW OF EXPERIMENTAL CAPABILITIES AND HYDRODYNAMIC DATA FOR VALIDATION OF CFD-BASED PREDICTIONS FOR SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Daniel S. Wendt; Steven P. Antal; Michael Z. Podowski

    2007-11-01

    The purpose of this paper is to document the review of several open-literature sources of both experimental capabilities and published hydrodynamic data to aid in the validation of a Computational Fluid Dynamics (CFD) based model of a slurry bubble column (SBC). The review included searching the Web of Science, ISI Proceedings, and Inspec databases, internet searches as well as other open literature sources. The goal of this study was to identify available experimental facilities and relevant data. Integral (i.e., pertaining to the SBC system), as well as fundamental (i.e., separate effects are considered), data are included in the scope of this effort. The fundamental data is needed to validate the individual mechanistic models or closure laws used in a Computational Multiphase Fluid Dynamics (CMFD) simulation of a SBC. The fundamental data is generally focused on simple geometries (i.e., flow between parallel plates or cylindrical pipes) or custom-designed tests to focus on selected interfacial phenomena. Integral data covers the operation of a SBC as a system with coupled effects. This work highlights selected experimental capabilities and data for the purpose of SBC model validation, and is not meant to be an exhaustive summary.

  8. REVIEW OF EXPERIMENTAL CAPABILITIES AND HYDRODYNAMIC DATA FOR VALIDATION OF CFD BASED PREDICTIONS FOR SLURRY BUBBLE COLUMN REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Daniel S. Wendt

    2007-11-01

    The purpose of this paper is to document the review of several open-literature sources of both experimental capabilities and published hydrodynamic data to aid in the validation of a Computational Fluid Dynamics (CFD) based model of a slurry bubble column (SBC). The review included searching the Web of Science, ISI Proceedings, and Inspec databases, internet searches as well as other open literature sources. The goal of this study was to identify available experimental facilities and relevant data. Integral (i.e., pertaining to the SBC system), as well as fundamental (i.e., separate effects are considered), data are included in the scope of this effort. The fundamental data is needed to validate the individual mechanistic models or closure laws used in a Computational Multiphase Fluid Dynamics (CMFD) simulation of a SBC. The fundamental data is generally focused on simple geometries (i.e., flow between parallel plates or cylindrical pipes) or custom-designed tests to focus on selected interfacial phenomena. Integral data covers the operation of a SBC as a system with coupled effects. This work highlights selected experimental capabilities and data for the purpose of SBC model validation, and is not meant to be an exhaustive summary.

  9. Final report on in-reactor creep-fatigue deformation behaviour of a CuCrZr alloy: COFAT 1

    DEFF Research Database (Denmark)

    Singh, Bachu Narain; Tähtinen, S.; Moilanen, P.

    CrZr(HT1) alloy exposed concurrently to flux of neutrons and creep-fatigue cyclic loading directly in a fission reactor. Special experimental facilities were designed and fabricated for this purpose. A number of in-reactor creep-fatigue experiments were successfully carried out in the BR-2 reactor at Mol...... (Belgium). In the present report we first describe the experimental facilities and the details of the in-reactor creep-fatigue experiments carried out at 363 and 343K at a strain amplitude of 0.5% with holdtimes of 10 and 100s, respectively. For comparison purposes, similar creep-fatigue tests were...... performed outside of the reactor. (i.e. in the absence of neutron irradiation). During in-reactor tests, the mechanical response was continuously registered throughout the whole test. The results are first presented in the form of hysteresis loops confirming that the nature of deformation during these tests...

  10. Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick Nathan [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, Mark David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

  11. The economic and community impacts of closing Hanford's N Reactor and nuclear materials production facilities

    Energy Technology Data Exchange (ETDEWEB)

    Scott, M.J.; Belzer, D.B.; Nesse, R.J.; Schultz, R.W.; Stokowski, P.A.; Clark, D.C.

    1987-08-01

    This study discusses the negative economic impact on local cities and counties and the State of Washington of a permanent closure of nuclear materials production at the Hanford Site, located in the southeastern part of the state. The loss of nuclear materials production, the largest and most important of the five Department of Energy (DOE) missions at Hanford, could occur if Hanford's N Reactor is permanently closed and not replaced. The study provides estimates of statewide and local losses in jobs, income, and purchases from the private sector caused by such an event; it forecasts impacts on state and local government finances; and it describes certain local community and social impacts in the Tri-Cities (Richland, Kennewick, and Pasco) and surrounding communities. 33 refs., 8 figs., 22 tabs.

  12. Occupational radiation exposure at commercial nuclear power reactors and other facilities 1996: Twenty-ninth annual report. Volume 18

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, M.L. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications; Hagemeyer, D. [Science Applications International Corp., Oak Ridge, TN (United States)

    1998-02-01

    This report summarizes the occupational exposure data that are maintained in the US Nuclear Regulatory Commission`s (NRC) Radiation Exposure Information and Reporting System (REIRS). The bulk of the information contained in the report was compiled from the 1996 annual reports submitted by six of the seven categories of NRC licensees subject to the reporting requirements of 10 CFR 20.2206. Since there are no geologic repositories for high level waste currently licensed, only six categories will be considered in this report. Annual reports for 1996 were received from a total of 300 NRC licensees, of which 109 were operators of nuclear power reactors in commercial operation. Compilations of the reports submitted by the 300 licensees indicated that 138,310 individuals were monitored, 75,139 of whom received a measurable dose. The collective dose incurred by these individuals was 21,755 person-cSv (person-rem){sup 2} which represents a 13% decrease from the 1995 value. The number of workers receiving a measurable dose also decreased, resulting in the average measurable dose of 0.29 cSv (rem) for 1996. The average measurable dose is defined to be the total collective dose (TEDE) divided by the number of workers receiving a measurable dose. These figures have been adjusted to account for transient reactor workers. Analyses of transient worker data indicate that 22,348 individuals completed work assignments at two or more licensees during the monitoring year. The dose distributions are adjusted each year to account for the duplicate reporting of transient workers by multiple licensees. In 1996, the average measurable dose calculated from reported was 0.24 cSv (rem). The corrected dose distribution resulted in an average measurable dose of 0.29 cSv (rem).

  13. Reactor Safety Research Programs Quarterly Report July- September 1980

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1980-12-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  14. Reactor Safety Research Programs Quarterly Report April -June 1980

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1980-11-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from April 1 through June 30, 1980, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission {NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  15. Bioregenerative Life Support Experiment for 90-days in a Closed Integrative Experimental Facility LUNAR PALACE 1

    Science.gov (United States)

    Liu, Hong

    A 90-day bioregenerative life support experiment with three-member crew was carried out in the closed integrative experimental facility, LUNAR PALACE 1 regenerating basic living necessities and disposing wastes to provide life support for crew. It was composed of higher plant module, animal module, and waste treatment module. The higher plant module included wheat, chufa, pea, carrot and green leafy vegetables, with aim to satisfy requirement of 60% plant food and 100% O2 and water for crew. The yellow mealworm was selected as animal module to provide partial animal protein for crew, and reared on plant inedible biomass. The higher plant and yellow mealworm were both cultivated and harvested in the conveyor-type manner. The partial plant inedible biomass and human feces were mixed and co- fermented in the waste treatment module for preparation of soil-like substrate by bioconversion, maintaining gas balance and increasing closure degree. Meanwhile, in the waste treatment module, the water and partial nitrogen from human urine were recovered by physical-chemical means. Circulation of O2 and water as well as food supply from crops cultivated in the LUNAR PALACE 1 were investigated and calculated, and simultaneously gas exchange, mass flow among different components and system closure degree were also analyzed, respectively. Furthermore, the system robustness with respect to internal variation was tested and evaluated by sensitivity analysis of the aggregative index consisting of key performance indicators like crop yield, gaseous equilibrium concentration, microbial community composition, biogenic elements dynamics, etc., and comprehensively evaluating the operating state, to number change of crew from 2 to 4 during the 90-day closed experiment period.

  16. Commissioning of the STAR test section for experimental simulation of loss of coolant accident using the EC-208 instrumented fuel assembly of the IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maprelian, Eduardo; Torres, Walmir M.; Prado, Adelk C.; Umbehaun, Pedro E.; Franca, Renato L.; Santos, Samuel C.; Macedo, Luiz A.; Sabundjian, Gaiane, E-mail: emaprel@ipen.br, E-mail: wmtorres@ipen.br, E-mail: acprado@ipen.br, E-mail: umbehaun@ipen.br, E-mail: rlfranca@ipen.br, E-mail: samuelcs@ipen.br, E-mail: lamacedo@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SO (Brazil)

    2015-07-01

    The three basic safety functions of Research Reactors (RR) are the safe shutdown of the reactor, the proper cooling of the decay heat of the fuel elements and the confinement of radioactive materials. Compared to Nuclear Power Reactors, RR power release is small, yet its three safety functions must be met to ensure the integrity of the reactor. During a loss of coolant accident (LOCA) in pool type RR, partial or complete loss of pool water may occur, with consequent partial or complete uncovering of the fuel assemblies. In such an accident, the decay heat removal safety function must not be compromised. The Test Section for Experimental Simulation of Loss of Coolant Accident (STAR) is in commissioning phase. This test section will provide experimental data on partial and total uncovering of the EC-208 instrumented fuel assembly (IFA) irradiated in the IEA-R1. Experimental results will be useful in validation of computer codes for RR safety analysis, particularly on heat removal efficiency aspects (safety function) in accident conditions. STAR comprises a base on which is installed the IFA, the cylindrical stainless steel hull, the compressed air system for the test section emptying and refilling, and the instrumentation for temperature and level measurements. The commissioning tests or pre-operational check, consist of several preliminary tests to verify experimental procedures, the difficulties during assembling of STAR in the pool, the difficulties in control the emptying and refilling velocities, as well as, the repeatability capacity, tests of equipment, valves and systems and tests of instrumentation and data acquisition system. Safety, accuracy and easiness of operation will be checked. (author)

  17. Nitritation performance and biofilm development of co- and counter-diffusion biofilm reactors: Modeling and experimental comparison

    DEFF Research Database (Denmark)

    Wang, Rongchang; Terada, Akihiko; Lackner, Susanne

    2009-01-01

    A comparative study was conducted on the start-up performance and biofilm development in two different biofilm reactors with aim of obtaining partial nitritation. The reactors were both operated under oxygen limited conditions, but differed in geometry. While substrates (O-2, NH3) co...... results showed that the counter-diffusion biofilms developed faster and attained a larger maximum biofilm thickness than the co-diffusion biofilms. Under oxygen limited condition (DO

  18. Reactor Safety Research Programs Quarterly Report October - December 1981

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1982-03-01

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

  19. Reactor Safety Research Programs Quarterly Report July - September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S. K.

    1982-01-01

    This document summarizes the work performed by Pacific Northwest laboratory (PNL) from July 1 through September 30, 1981, for the Division of Accident Evaluation, U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR} steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipe-to-pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, lspra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL), Idaho Falls, Idaho. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  20. Experimental evaluation of two different types of reactors for CO2 removal from gaseous stream by bottom ash accelerated carbonation.

    Science.gov (United States)

    Lombardi, L; Carnevale, E A; Pecorini, I

    2016-12-01

    Low methane content landfill gas may be enriched by removing carbon dioxide. An innovative process, based on carbon dioxide capture and storage by means of accelerated carbonation of bottom ash is proposed and studied for the above purpose. Within this research framework we devoted a preliminary research activity to investigate the possibility of improving the way the contact between bottom ash and landfill gas takes place: this is the scope of the work reported in this paper. Two different types of reactors - fixed bed and rotating drum - were designed and constructed for this purpose. The process was investigated at laboratory scale. As the aim of this phase was the comparison of the performances of the two different reactors, we used a pure stream of CO2 to preliminarily evaluate the reactor behaviors in the most favorable condition for the process (i.e. maximum CO2 partial pressure at ambient condition). With respect to the simple fixed bed reactor concept, some modifications were proposed, consisting of separating the ash bed in three layers. With the three layer configuration we would like to reduce the possibility for the gas to follow preferential paths through the ash bed. However, the results showed that the process performances are not significantly influenced by the multiple layer arrangement. As an alternative to the fixed bed reactor, the rotating drum concept was selected in order to provide continuous mixing of the solids. Two operating parameters were considered and varied during the tests: the filling ratio and the rotating speed. Better performances were observed for lower filling ratio while the rotating speed showed minor importance. Finally the performances of the two reactors were compared. The rotating drum reactor is able to provide improved carbon dioxide removal with respect to the fixed bed one, especially when the rotating reactor is operated at low filling ratio values and slow rotating speed values. Comparing the carbon dioxide

  1. Systems Design and Experimental Evaluation of a High-Altitude Relight Test Facility

    Science.gov (United States)

    Paxton, Brendan

    Novel advances in gas turbine engine combustor technology, led by endeavors into fuel efficiency and demanding environmental regulations, have been fraught with performance and safety concerns. While the majority of low emissions gas turbine engine combustor technology has been necessary for power generation applications, the push for ultra-low NOx combustion in aircraft jet engines has been ever present. Recent state-of-the-art combustor designs notably tackle historic emissions challenges by operating at fuel-lean conditions, which are characterized by an increase in the amount of air flow sent to the primary combustion zone. While beneficial in reducing NOx emissions, the fuel-lean mechanisms that characterize these combustor designs rely heavily upon high-energy and high-velocity air flows to sufficiently mix and atomize fuel droplets, ultimately leading to flame stability concerns during low-power operation. When operating at high-altitude conditions, these issues are further exacerbated by the presence of low ambient air pressures and temperatures, which can lead to engine flame-out situations and hamper engine relight attempts. To aid academic and industrial research ventures into improving the high-altitude lean blow-out and relight performance of modern gas turbine engine combustor technologies, the High-Altitude Relight Test Facility (HARTF) was designed and constructed at the University of Cincinnati (UC) Combustion and Fire Research Laboratory (CFRL). Following its construction, an experimental evaluation of its abilities to facilitate optically-accessible ignition, combustion, and spray testing for gas turbine engine combustor hardware at simulated high-altitude conditions was performed. In its evaluation, performance limit references were established through testing of the HARTF vacuum and cryogenic air-chilling capabilities. These tests were conducted with regard to end-user control---the creation and the maintenance of a realistic high

  2. Pressurizing new reactors

    Energy Technology Data Exchange (ETDEWEB)

    Neill, J.S.

    1956-01-30

    The Technical Division was asked recently to consider designs for new reactors that would add 8000 MW capacity to the Savannah River Plant. One modification of the existing SRP design that would enable a higher power rating, and therefore require fewer new reactors, is an increase in the maximum pressure in the D{sub 2}O system. The existing reactors at SRP are designed for a maximum pressure in the gas plenum of only 5 psig. Higher pressures enable higher D{sub 2} temperatures and higher sheath temperatures without local boiling or burnout. The requirements in reactor cooling facilities at any given power level would therefore be reduced by pressurizing.

  3. A method for development of efficient 3D models for neutronic calculations of ASTRA critical facility using experimental information

    Energy Technology Data Exchange (ETDEWEB)

    Balanin, A. L.; Boyarinov, V. F.; Glushkov, E. S.; Zimin, A. A.; Kompaniets, G. V.; Nevinitsa, V. A., E-mail: Neviniza-VA@nrcki.ru; Moroz, N. P.; Fomichenko, P. A.; Timoshinov, A. V. [National Research Center Kurchatov Institute (Russian Federation); Volkov, Yu. N. [National Research Nuclear University MEPhI (Russian Federation)

    2016-12-15

    The application of experimental information on measured axial distributions of fission reaction rates for development of 3D numerical models of the ASTRA critical facility taking into account azimuthal asymmetry of the assembly simulating a HTGR with annular core is substantiated. Owing to the presence of the bottom reflector and the absence of the top reflector, the application of 2D models based on experimentally determined buckling is impossible for calculation of critical assemblies of the ASTRA facility; therefore, an alternative approach based on the application of the extrapolated assembly height is proposed. This approach is exemplified by the numerical analysis of experiments on measurement of efficiency of control rods mockups and protection system (CPS).

  4. Earth Systems Questions in Experimental Climate Change Science: Pressing Questions and Necessary Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Osmond, B.

    2002-05-20

    Sixty-four scientists from universities, national laboratories, and other research institutions worldwide met to evaluate the feasibility and potential of the Biosphere2 Laboratory (B2L) as an inclusive multi-user scientific facility (i.e., a facility open to researchers from all institutions, according to agreed principles of access) for earth system studies and engineering research, education, and training relevant to the mission of the United States Department of Energy (DOE).

  5. Dr Robert Aymar, Director of the International Thermonuclear Experimental Reactor (ITER), was nominated to succeed Professor Luciano Maiani as CERN's Director General, to take office on 1 January 2004.

    CERN Document Server

    2002-01-01

    Dr Robert Aymar, Director of the International Thermonuclear Experimental Reactor (ITER), was nominated to succeed Professor Luciano Maiani as CERN's Director General, to take office on 1 January 2004.

  6. Experimental study of local coolant hydrodynamics in TVS-Kvadrat PWR reactor fuel assembly using mixing spacer grids with different types of deflectors

    Directory of Open Access Journals (Sweden)

    S.M. Dmitriev

    2015-12-01

    Full Text Available Results of experimental studies of local hydrodynamic characteristics of coolant flow in fuel assemblies of RWR reactors using different types of mixing spacer grids are presented. Specific features and regularities of coolant flow in fuel pin bundles of TVS-KVADRAT fuel assemblies with different types of mixing spacer grids were revealed in the course of experiments. Analysis of space distribution of projections of absolute flow velocity allowed detailed description of coolant flow beyond the spacer grid with installation of three different types of deflectors. Optimal design of deflector for spacer grid of the TVS-KVADRAT fuel assembly in the standard cell in the area of guiding channels was identified. Results of studies of local hydrodynamics of coolant flow in the TVS-KVADRAT fuel assembly are accepted for subsequent practical application by the JSC Afrikantov Experimental Design Bureau for Mechanical Engineering (OKBM in the evaluations of thermal engineering reliability of PWR reactor cores and were included in the database for verification of computational fluid dynamic codes (CFD-codes and implementation of detailed cell array calculations of PWR reactor cores.

  7. Development of essential system technologies for advanced reactor - Development of natural circulation analysis code for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Park, Ik Gyu; Kim, Jae Hak; Lee, Sang Min; Kim, Tae Wan [Seoul National University, Seoul (Korea)

    1999-04-01

    The objective of this study is to understand the natural circulation characteristics of integral type reactors and to develope the natural circulation analysis code for integral type reactors. This study is focused on the asymmetric 3-dimensional flow during natural circulation such as 1/4 steam generator section isolation and the inclination of the reactor systems. Natural circulation experiments were done using small-scale facilities of integral reactor SMART (System-Integrated Modular Advanced ReacTor). CFX4 code was used to investigate the flow patterns and thermal mixing phenomena in upper pressure header and downcomer. Differences between normal operation of all steam generators and the 1/4 section isolation conditions were observed and the results were used as the data 1/4 section isolation conditions were observed and the results were used as the data for RETRAN-03/INT code validation. RETRAN-03 code was modified for the development of natural circulation analysis code for integral type reactors, which was development of natural circulation analysis code for integral type reactors, which was named as RETRAN-03/INT. 3-dimensional analysis models for asymmetric flow in integral type reactors were developed using vector momentum equations in RETRAN-03. Analysis results using RETRAN-03/INT were compared with experimental and CFX4 analysis results and showed good agreements. The natural circulation characteristics obtained in this study will provide the important and fundamental design features for the future small and medium integral reactors. (author). 29 refs., 75 figs., 18 tabs.

  8. Reactor operation safety information document

    Energy Technology Data Exchange (ETDEWEB)

    1990-01-01

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  9. Wind refrigeration : design and results of an experimental facility; Refrigeracion eolica: Diseno y resultados de una instalacion experimental

    Energy Technology Data Exchange (ETDEWEB)

    Beltran, R. G.; Talero, A.

    2004-07-01

    This article describes the experimental setup used to obtain design parameters for a wind driven refrigeration equipment. The system compressor is directly coupled to the wind mill and will provide refrigeration to a community located in La Guajira in northern Colombia. The testing on the experimental installation assessed the refrigeration capacity that could be provided by an open type commercial compressor coupled to the wind mill axis. Power and torque requirements have been evaluated for different wind mill rotational speeds. An assessment of the local conditions relating to wind speed, frequency and preferred direction for the installation site has been made based on measurements by the Meteorological National Institute and independent data from other sources. (Author)

  10. Use the results of measurements on KBR facility for testing of neutron data of main structural materials for fast reactors

    Science.gov (United States)

    Koscheev, Vladimir; Manturov, Gennady; Pronyaev, Vladimir; Rozhikhin, Evgeny; Semenov, Mikhail; Tsibulya, Anatoly

    2017-09-01

    Several k∞ experiments were performed on the KBR critical facility at the Institute of Physics and Power Engineering (IPPE), Obninsk, Russia during the 1970s and 80s for study of neutron absorption properties of Cr, Mn, Fe, Ni, Zr, and Mo. Calculations of these benchmarks with almost any modern evaluated nuclear data libraries demonstrate bad agreement with the experiment. Neutron capture cross sections of the odd isotopes of Cr, Mn, Fe, and Ni in the ROSFOND-2010 library have been reevaluated and another evaluation of the Zr nuclear data has been adopted. Use of the modified nuclear data for Cr, Mn, Fe, Ni, and Zr leads to significant improvement of the C/E ratio for the KBR assemblies. Also a significant improvement in agreement between calculated and evaluated values for benchmarks with Fe reflectors was observed. C/E results obtained with the modified ROSFOND library for complex benchmark models that are highly sensitive to the cross sections of structural materials are no worse than results obtained with other major evaluated data libraries. Possible improvement in results by decreasing the capture cross section for Zr and Mo at the energies above 1 keV is indicated.

  11. Use the results of measurements on KBR facility for testing of neutron data of main structural materials for fast reactors

    Directory of Open Access Journals (Sweden)

    Koscheev Vladimir

    2017-01-01

    Full Text Available Several k∞ experiments were performed on the KBR critical facility at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia during the 1970s and 80s for study of neutron absorption properties of Cr, Mn, Fe, Ni, Zr, and Mo. Calculations of these benchmarks with almost any modern evaluated nuclear data libraries demonstrate bad agreement with the experiment. Neutron capture cross sections of the odd isotopes of Cr, Mn, Fe, and Ni in the ROSFOND-2010 library have been reevaluated and another evaluation of the Zr nuclear data has been adopted. Use of the modified nuclear data for Cr, Mn, Fe, Ni, and Zr leads to significant improvement of the C/E ratio for the KBR assemblies. Also a significant improvement in agreement between calculated and evaluated values for benchmarks with Fe reflectors was observed. C/E results obtained with the modified ROSFOND library for complex benchmark models that are highly sensitive to the cross sections of structural materials are no worse than results obtained with other major evaluated data libraries. Possible improvement in results by decreasing the capture cross section for Zr and Mo at the energies above 1 keV is indicated.

  12. Novel test-bed facility for PSI issues in fusion reactor conditions on the base of next generation QSPA plasma accelerator

    Science.gov (United States)

    Garkusha, I. E.; Chebotarev, V. V.; Herashchenko, S. S.; Makhlaj, V. A.; Kulik, N. V.; Ladygina, M. S.; Marchenko, A. K.; Petrov, Yu. V.; Staltsov, V. V.; Shevchuk, P. V.; Solyakov, D. G.; Yelisyeyev, D. V.

    2017-11-01

    In this report a concept of a new generation QSPA with external B-field up to 2 T has been discussed. A novel test-bed facility, which was recently constructed in Kharkov IPP NSC KIPT, has been described. It allows for a new level of plasma stream parameters and its wide variation in new QSPA-M device, as well as possible combination of steady-state and pulsed plasma loads to the materials during the exposures. First plasma is recently obtained. Careful optimization of the operational regimes of the plasma accelerator’s functional components and plasma dynamics in the magnetic system of QSPA-M device has started approaching step by step the necessary level of plasma parameters and their effective variation. The relevant results on plasma stream characterization are presented. Energy density distributions in plasma stream have been measured with calorimetry. Spectroscopy and probe technique have also been applied for plasma parameters measurements. The obtained results demonstrate the ability of QSPA-M to reproduce the ELM impacts in fusion reactor, both in terms of heat load and particle flux to the surface.

  13. High-field conductor testing at the FENIX facility

    Science.gov (United States)

    Shen, S. S.; Chaplin, M. R.; Felker, B.; Hassenzahl, W. V.; Kishiyama, K. I.; Parker, J. M.

    1993-04-01

    The Fusion Engineering International experiments (FENIX) Test Facility, which was commissioned at the end of 1991, is the first facility in the world capable of testing prototype conductors for the International Thermonuclear Experimental Reactor (ITER) superconducting magnets. The FENIX facility provides test conditions that simulate the ITER magnet operating environment; more importantly, it also accommodates specific experiments to determine the operational margins for the prototype conductors. The FENIX facility generates magnetic fields close to 14 T, and transport currents over 40 kA for testing the prototype conductors. This paper describes an experimental program that measures critical currents, current-sharing temperatures, forced-flow properties, and cyclic effects.

  14. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  15. Autothermal reforming of methane with integrated CO2 capture in a novel fluidized bed membrane reactor. Part 1: experimental demonstration

    NARCIS (Netherlands)

    Gallucci, F.; van Sint Annaland, M.; Kuipers, J.A.M.

    2008-01-01

    Two fluidized bed membrane reactor concepts for hydrogen production via autothermal reforming of methane with integrated CO2 capture are proposed. Ultra-pure hydrogen is obtained via hydrogen perm-selective Pd-based membranes, while the required reaction energy is supplied by oxidizing part of the

  16. Experimental study of a cocurrent upflow packed bed bubble column reactor: pressure drop, holdup and interfacial area

    NARCIS (Netherlands)

    Molga, E.J.; Westerterp, K.R.

    1997-01-01

    Gas¿liquid interfacial areas have been determined by means of chemically enhanced absorption of CO2 into DEA in a packed bed bubble column reactor with an inner diameter of 156 mm. The influence of the gas velocity and particle diameter on the interfacial areas, pressure drops and liquid holdups has

  17. Fluidised bed membrane reactor for ultrapure hydrogen production via methane steam reforming: Experimental demonstration and model validation

    NARCIS (Netherlands)

    Patil, C.S.; van Sint Annaland, M.; Kuipers, J.A.M.

    2007-01-01

    Hydrogen is emerging as a future alternative for mobile and stationary energy carriers in addition to its use in chemical and petrochemical applications. A novel multifunctional reactor concept has been developed for the production of ultrapure hydrogen View the MathML source from light hydrocarbons

  18. Reactor Safety Research Programs Quarterly Report October - December 1980

    Energy Technology Data Exchange (ETDEWEB)

    Edler, S K

    1981-04-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from October 1 through December 31, 1980, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining structural graphite strength, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NOE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-of-coolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation and postaccident coolability tests for the ESSOR Test Reactor Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  19. Safety of magnetic fusion facilities: Volume 2, Guidance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    This document provides guidance for the implementation of the requirements identified in Vol. 1 of this Standard. This guidance is intended for the managers, designers, operators, and other personnel with safety responsibilities for facilities designated as magnetic fusion facilities. While Vol. 1 is generally applicable in that requirements there apply to a wide range of fusion facilities, this volume is concerned mainly with large facilities such as the International Thermonuclear Experimental Reactor (ITER). Using a risk-based prioritization, the concepts presented here may also be applied to other magnetic fusion facilities. This volume is oriented toward regulation in the Department of Energy (DOE) environment.

  20. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  1. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  2. Experimental investigation and SIMMER-III code modelling of LBE–water interaction in LIFUS5/Mod2 facility

    Energy Technology Data Exchange (ETDEWEB)

    Pesetti, A., E-mail: alessio.pesetti@for.unipi.it [University of Pisa, DICI, 56122 Pisa, PI (Italy); Del Nevo, A. [ENEA, C.R. Brasimone, 40032 Camugnano, BO (Italy); Forgione, N. [University of Pisa, DICI, 56122 Pisa, PI (Italy)

    2015-08-15

    Highlights: • In the framework of THINS Project, four experimental tests were performed investigating water–LBE interaction in LIFUS5/Mod2 facility at ENEA RC Brasimone. • High quality data were acquired to improve the phenomena understanding and support codes development/validation. • Post-test activity by SIMMER III code was performed by a simplified model. • The obtained numerical results are qualitatively in agreement with the experimental data. - Abstract: In the frame of the THINS Project, an experimental campaign was performed on LIFUS5/Mod2 facility at ENEA RC Brasimone, aiming to investigate the water–LBE interaction. Such a phenomenon occurs as consequence of a postulated Steam Generator Tube Rupture event in a HLMFR system. Four tests were performed injecting sub-cooled water at 40 bar into a reaction vessel partially filled by low pressure LBE at 400 °C. The post-test activity was performed by the SIMMER-III code in order to improve the understanding of the involved phenomena and to confirm the code capabilities in simulating the water–LBE interaction. The calculated data showed a qualitatively agreement with the measured values and a faster reaction kinetics due to the modelling assumptions.

  3. Validation of raw experimental data during shoting at the LIL facility

    Science.gov (United States)

    Henry, Olivier; Domin, Vincent; Romary, Philippe; Raffestin, Didier

    2012-10-01

    The LIL (Laser Integration Line) facility at CESTA (Aquitaine, France) is a facility allowing the delivery of 20 kJ at 3φ. The experiment system includes 13 diagnostics. The facility must be able to deliver, within one hour following shoting, all the results of the plasma diagnostics, alignment images and laser diagnostic measurements. These results have to be guaranteed in terms of conformity to the request and quality of measurement. The LIL has developed a tool for the visualisation, analysis and validation of the data. The software is written in the Delphi language for the main body. The configuration is based on XML files. It is thus possible to re-read the external analysis modules in Python (the language used on the future LMJ). The software is built on three pillars: definition of a validation model prior to the campaign, basic physical models to qualify the signal as compliant and exploitable, and inter-comparison of the shoting and signals over a given campaign or period. Validation of the raw plasma data must serve to validate and guarantee performances, assure the conformity of the PD configuration to the request from the client, check the consistency of measurements, trigger corrective maintenance if necessary.

  4. The ΣΣ secondary intermediate-energy standard neutron field development at the Romania TRIGA Reactor

    Science.gov (United States)

    Roth, C.; Gârlea, I.; Dobrea, D.; Aioanei, L.; Kelerman, C.; Gârlea, C.; Gugiu, D.; Datcu, A.; Preda, M.; Pavelescu, M.

    2004-04-01

    The ΣΣ intermediate-energy reference spectrum irradiation facility, operated until 1998 at the VVR-S Reactor from Bucharest, was put into operation at the Romania TRIGA Reactor. This paper presents the experimental devices developed for the ΣΣ system operating at the Steady State Reactor (SSR)-TRIGA core, including the monitoring system, and preliminary neutron characterization results. These results show that the spectrum characteristics are slowly deviating from those for the recommended ΣΣ spectrum. The high-energy neutron tail is slowly increasing, compared with the recommended ΣΣ spectrum and other similar facilities.

  5. Nuclear reactors built, being built, or planned, 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

    This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1994. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; tables of data for reactors operating, being built, or planned; and tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company -- working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  6. Nuclear reactors built, being built, or planned: 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    This report contains unclassified information about facilities built, being built, or planned in the US for domestic use or export as of December 31, 1995. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE; from the US Nuclear Regulatory Commission (NRC); from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The book consists of three divisions, as follows: (1) a commercial reactor locator map and tables of the characteristic and statistical data that follow; a table of abbreviations; (2) tables of data for reactors operating, being built, or planned; and (3) tables of data for reactors that have been shut down permanently or dismantled. The reactors are subdivided into the following parts: Civilian, Production, Military, Export, and Critical Assembly. Export reactor refers to a reactor for which the principal nuclear contractor is a US company--working either independently or in cooperation with a foreign company (Part 4). Critical assembly refers to an assembly of fuel and moderator that requires an external source of neutrons to initiate and maintain fission. A critical assembly is used for experimental measurements (Part 5).

  7. Oxidative coupling of methane using inorganic membrane reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Y.H.; Moser, W.R.; Dixon, A.G. [Worcester Polytechnic Institute, MA (United States)] [and others

    1995-12-31

    The goal of this research is to improve the oxidative coupling of methane in a catalytic inorganic membrane reactor. A specific target is to achieve conversion of methane to C{sub 2} hydrocarbons at very high selectivity and relatively higher yields than in fixed bed reactors by controlling the oxygen supply through the membrane. A membrane reactor has the advantage of precisely controlling the rate of delivery of oxygen to the catalyst. This facility permits balancing the rate of oxidation and reduction of the catalyst. In addition, membrane reactors minimize the concentration of gas phase oxygen thus reducing non selective gas phase reactions, which are believed to be a main route for formation of CO{sub x} products. Such gas phase reactions are a cause for decreased selectivity in oxidative coupling of methane in conventional flow reactors. Membrane reactors could also produce higher product yields by providing better distribution of the reactant gases over the catalyst than the conventional plug flow reactors. Modeling work which aimed at predicting the observed experimental trends in porous membrane reactors was also undertaken in this research program.

  8. Experimental determination of spectral indices by scanning of fuel rod in the IPEN/MB-01 reactor; Determinacao experimental de indices espectrais por varredura gama de vareta combustivel no reator IPEN/MB-01

    Energy Technology Data Exchange (ETDEWEB)

    Fanaro, Leda Cristina Cabelo Bernardes

    2009-07-01

    In this work, the spectral indexes 28{rho}{sup *} and 25{delta}{sup *}, and gamma efficiency factor in the IPEN/MB-01 reactor were determined experimentally employing a rod scanning technique. In the case of 28{rho}{sup *}, this method has the advantage of eliminating most of the correction factors derived from the calculations. Only the fission yield factor and the relative fission rate in the {sup 235}U remain in the determination of the 25{delta}{sup *}. The experiments were performed with different thicknesses of cadmium sleeves: 0.55 mm, 1.10 mm and 2.20 mm. The final experimental uncertainty achieved in the experiment, less that 1 %, and the excellent geometrical and material data characterization of the IPEN/MB-01 reactor allow to use the results as benchmark for validate calculation methods and related nuclear data libraries. The comparison between calculated values and experimental values was performed by employing the MCNP-5 code and the nuclear data libraries: ENDF/B-VI.8, ENDF/B-VII.0, JENDL-3.3 and JEFF-3.1. The results demonstrate that the difference among libraries is very small. Also, the comparison between calculated values and experimental values shows that there has been considerable progress in the recent nuclear data libraries. The best result is obtained with ENDF/B-VII.0 nuclear data library, and the highest discrepancy was obtained with JEFF-3.1 and JENDL-3.3 nuclear data libraries. (author)

  9. Energy efficient electrocoagulation using a new flow column reactor to remove nitrate from drinking water - Experimental, statistical, and economic approach.

    Science.gov (United States)

    Hashim, Khalid S; Shaw, Andy; Al Khaddar, Rafid; Pedrola, Montserrat Ortoneda; Phipps, David

    2017-07-01

    In this investigation, a new bench-scale electrocoagulation reactor (FCER) has been applied for drinking water denitrification. FCER utilises the concepts of flow column to mix and aerate the water. The water being treated flows through the perforated aluminium disks electrodes, thereby efficiently mixing and aerating the water. As a result, FCER reduces the need for external stirring and aerating devices, which until now have been widely used in the electrocoagulation reactors. Therefore, FCER could be a promising cost-effective alternative to the traditional lab-scale EC reactors. A comprehensive study has been commenced to investigate the performance of the new reactor. This includes the application of FCER to remove nitrate from drinking water. Estimation of the produced amount of H2 gas and the yieldable energy from it, an estimation of its preliminary operating cost, and a SEM (scanning electron microscope) investigation of the influence of the EC process on the morphology of the surface of electrodes. Additionally, an empirical model was developed to reproduce the nitrate removal performance of the FCER. The results obtained indicated that the FCER reduced the nitrate concentration from 100 to 15 mg/L (World Health Organization limitations for infants) after 55 min of electrolysing at initial pH of 7, GBE of 5 mm, CD of 2 mA/cm2, and at operating cost of 0.455 US $/m3. Additionally, it was found that FCER emits H2 gas enough to generate a power of 1.36 kW/m3. Statistically, the relationship between the operating parameters and nitrate removal could be modelled with R2 of 0.848. The obtained SEM images showed a large number dents on anode's surface due to the production of aluminium hydroxides. Crown Copyright © 2017. Published by Elsevier Ltd. All rights reserved.

  10. The Integral Test Facility Karlstein

    Directory of Open Access Journals (Sweden)

    Stephan Leyer

    2012-01-01

    Full Text Available The Integral Test Facility Karlstein (INKA test facility was designed and erected to test the performance of the passive safety systems of KERENA, the new AREVA Boiling Water Reactor design. The experimental program included single component/system tests of the Emergency Condenser, the Containment Cooling Condenser and the Passive Core Flooding System. Integral system tests, including also the Passive Pressure Pulse Transmitter, will be performed to simulate transients and Loss of Coolant Accident scenarios at the test facility. The INKA test facility represents the KERENA Containment with a volume scaling of 1 : 24. Component heights and levels are in full scale. The reactor pressure vessel is simulated by the accumulator vessel of the large valve test facility of Karlstein—a vessel with a design pressure of 11 MPa and a storage capacity of 125 m3. The vessel is fed by a benson boiler with a maximum power supply of 22 MW. The INKA multi compartment pressure suppression Containment meets the requirements of modern and existing BWR designs. As a result of the large power supply at the facility, INKA is capable of simulating various accident scenarios, including a full train of passive systems, starting with the initiating event—for example pipe rupture.

  11. Modeling of the charge-state separation at ITEP experimental facility for material science based on a Bernas ion source.

    Science.gov (United States)

    Barminova, H Y; Saratovskyh, M S

    2016-02-01

    The experiment automation system is supposed to be developed for experimental facility for material science at ITEP, based on a Bernas ion source. The program CAMFT is assumed to be involved into the program of the experiment automation. CAMFT is developed to simulate the intense charged particle bunch motion in the external magnetic fields with arbitrary geometry by means of the accurate solution of the particle motion equation. Program allows the consideration of the bunch intensity up to 10(10) ppb. Preliminary calculations are performed at ITEP supercomputer. The results of the simulation of the beam pre-acceleration and following turn in magnetic field are presented for different initial conditions.

  12. Analytical study with the Athena code of the HCPB blanket design experimental activity on the HE-FUS3 facility

    Energy Technology Data Exchange (ETDEWEB)

    Meloni, P. [ENEA, Bologna (Italy); D' Auria, F.; Oriolo, F. [Pisa Univ. (Italy). Dipt. di Costruzioni Meccaniche e Nucleari; Dell' Orco, G.; Polazzi, G. [ENEA, Brasimone (Italy)

    1998-07-01

    Within the frame of the European Fusion Technology Programme, the HE-FUS3 helium facility was selected for the execution of thermal-hydraulic experimental campaigns on the HCPB Blanket mock-ups. The studies for the DEMO HCPB Blanket design, for the 1997-1998, contemplates a task concerning the Out-of-Pile tests for the HCPB Test Blanket Module (TBM). One of the main objectives of the task is the qualification of the HE-FUS3 capability to perform tests both in normal and off-normal conditions. (authors)

  13. Tests of the space gamma spectrometer prototype at the JINR experimental facility with different types of neutron generators

    Science.gov (United States)

    Litvak, M. L.; Vostrukhin, A. A.; Golovin, D. V.; Dubasov, P. V.; Zontikov, A. O.; Kozyrev, A. S.; Krylov, A. R.; Krylov, V. A.; Mitrofanov, I. G.; Mokrousov, M. I.; Repkin, A. N.; Timoshenko, G. N.; Udovichenko, K. V.; Shvetsov, V. N.

    2017-07-01

    The results of the tests of the HPGe gamma spectrometer performed with a planetary soil model and different types of pulse neutron generators are presented. All measurements have been performed at the experimental nuclear planetary science facility (Joint Institute for Nuclear Research) for the physical calibration of active gamma and neutron spectrometers. The aim of the study is to model a space experiment on determining the elemental composition of Martian planetary matter by neutron-induced gamma spectroscopy. The advantages and disadvantages of a gas-filled neutron generator in comparison with a vacuum-tube neutron generator are examined.

  14. Flow Reactor Studies with Nanosecond Pulsed Discharges at Atmospheric Pressure and Higher

    Science.gov (United States)

    2013-10-01

    Reactor Experiments: • Experiments on the oxidation of C1-C7 alkanes with and without plasma assisted reaction performed for T = 300 – 1250 K and P = 1...online) and Gas Chromatography (offline) Inlet Outlet Experimental Facility – cont’d • FTIR (online) connected to exhaust stream of reactor...molecule/pulse Major Species + Minor Species Oxygen + Excited Species C1-C7 Alkanes Plasma Assisted Oxidation Fourth Annual Review Meeting of

  15. The reactor core TRIGA Mark-III with fuels type 30/20; El nucleo del reactor TRIGA Mark-III con combustible tipo 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F., E-mail: fortunato.aguilar@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    This work describes the calculation series carried out with the program MCNP5 in order to define the configuration of the reactor core with fuels 30/20 (fuels with 30% of uranium content in the Or-Zr-H mixture and a nominal enrichment of 20%). To select the configuration of the reactor core more appropriate to the necessities and future uses of the reactor, the following criterions were taken into account: a) the excess in the reactor reactivity, b) the switch out margin and c) to have new irradiation facilities inside the reactor core. Taking into account these criterions is proceeded to know the characteristics of the components that form the reactor core (dimensions, geometry, materials, densities and positions), was elaborated a base model of the reactor core, for the MCNP5 code, with a configuration composed by 85 fuel elements, 4 control bars and the corresponding structural elements. The high reactivity excess obtained with this model, gave the rule to realize other models of the reactor core in which the reactivity excess and the switch out margin were approximate to the values established in the technical specifications of the reactor operation. Several models were realized until finding the satisfactory model; this is composite for 74 fuels, 4 control bars and 6 additional experimental positions inside the reactor core. (Author)

  16. The improvement of control rod in experimental fast reactor JOYO. The development of a sodium bonded type control rod

    Energy Technology Data Exchange (ETDEWEB)

    Soga, T.; Miyakawa, S.; Mitsugi, T. [Japan Nuclear Cycle Development Inst., Oarai Engineering Center, Irradiation Center, Irradiation and Administration Section, Oarai, Ibaraki (Japan)

    1999-06-01

    that the cladding strength sufficient to withstand stress accounting for decreased thickness by the ACCI zone. (5) The wet wash and storage method was selected for disposing of the spent sodium bonded control rods, based upon experimental results at the JOYO facilities. The effects from storing sodium bonded control rods in wet storage were evaluated. The results indicated that these effect would not pose a safety problem. (author)

  17. Remote Reactor Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Bernstein, Adam [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dazeley, Steve [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dobie, Doug [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Marleau, Peter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Brennan, Jim [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gerling, Mark [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sumner, Matthew [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sweany, Melinda [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2014-10-21

    The overall goal of the WATCHMAN project is to experimentally demonstrate the potential of water Cerenkov antineutrino detectors as a tool for remote monitoring of nuclear reactors. In particular, the project seeks to field a large prototype gadolinium-doped, water-based antineutrino detector to demonstrate sensitivity to a power reactor at ~10 kilometer standoff using a kiloton scale detector. The technology under development, when fully realized at large scale, could provide remote near-real-time information about reactor existence and operational status for small operating nuclear reactors out to distances of many hundreds of kilometers.

  18. Determination of the Clean Air Delivery Rate (CADR) of Photocatalytic Oxidation (PCO) Purifiers for Indoor Air Pollutants Using a Closed-Loop Reactor. Part II: Experimental Results.

    Science.gov (United States)

    Héquet, Valérie; Batault, Frédéric; Raillard, Cécile; Thévenet, Frédéric; Le Coq, Laurence; Dumont, Éric

    2017-03-06

    The performances of a laboratory PhotoCatalytic Oxidation (PCO) device were determined using a recirculation closed-loop pilot reactor. The closed-loop system was modeled by associating equations related to two ideal reactors: a perfectly mixed reservoir with a volume of VR = 0.42 m³ and a plug flow system corresponding to the PCO device with a volume of VP = 5.6 × 10-3 m³. The PCO device was composed of a pleated photocatalytic filter (1100 cm²) and two 18-W UVA fluorescent tubes. The Clean Air Delivery Rate (CADR) of the apparatus was measured under different operating conditions. The influence of three operating parameters was investigated: (i) light irradiance I from 0.10 to 2.0 mW·cm-2; (ii) air velocity v from 0.2 to 1.9 m·s-1; and (iii) initial toluene concentration C₀ (200, 600, 1000 and 4700 ppbv). The results showed that the conditions needed to apply a first-order decay model to the experimental data (described in Part I) were fulfilled. The CADR values, ranging from 0.35 to 3.95 m³·h-1, were mainly dependent on the light irradiance intensity. A square root influence of the light irradiance was observed. Although the CADR of the PCO device inserted in the closed-loop reactor did not theoretically depend on the flow rate (see Part I), the experimental results did not enable the confirmation of this prediction. The initial concentration was also a parameter influencing the CADR, as well as the toluene degradation rate. The maximum degradation rate rmax ranged from 342 to 4894 ppbv/h. Finally, this study evidenced that a recirculation closed-loop pilot could be used to develop a reliable standard test method to assess the effectiveness of PCO devices.

  19. Determination of the Clean Air Delivery Rate (CADR of Photocatalytic Oxidation (PCO Purifiers for Indoor Air Pollutants Using a Closed-Loop Reactor. Part II: Experimental Results

    Directory of Open Access Journals (Sweden)

    Valérie Héquet

    2017-03-01

    Full Text Available The performances of a laboratory PhotoCatalytic Oxidation (PCO device were determined using a recirculation closed-loop pilot reactor. The closed-loop system was modeled by associating equations related to two ideal reactors: a perfectly mixed reservoir with a volume of VR = 0.42 m3 and a plug flow system corresponding to the PCO device with a volume of VP = 5.6 × 10−3 m3. The PCO device was composed of a pleated photocatalytic filter (1100 cm2 and two 18-W UVA fluorescent tubes. The Clean Air Delivery Rate (CADR of the apparatus was measured under different operating conditions. The influence of three operating parameters was investigated: (i light irradiance I from 0.10 to 2.0 mW·cm−2; (ii air velocity v from 0.2 to 1.9 m·s−1; and (iii initial toluene concentration C0 (200, 600, 1000 and 4700 ppbv. The results showed that the conditions needed to apply a first-order decay model to the experimental data (described in Part I were fulfilled. The CADR values, ranging from 0.35 to 3.95 m3·h−1, were mainly dependent on the light irradiance intensity. A square root influence of the light irradiance was observed. Although the CADR of the PCO device inserted in the closed-loop reactor did not theoretically depend on the flow rate (see Part I, the experimental results did not enable the confirmation of this prediction. The initial concentration was also a parameter influencing the CADR, as well as the toluene degradation rate. The maximum degradation rate rmax ranged from 342 to 4894 ppbv/h. Finally, this study evidenced that a recirculation closed-loop pilot could be used to develop a reliable standard test method to assess the effectiveness of PCO devices.

  20. Initial design for an experimental investigation of strongly coupled plasma behavior in the ATLAS facility

    CERN Document Server

    Munson, C P; Taylor, A J; Trainor, R J; Wood, B P; Wysocki, F J

    1999-01-01

    Summary form only given. Atlas is a high current (~30 MA peak, with a current risetime ~4.5 mu sec), high energy (E/sub stored/=24 MJ, E /sub load/=3-6 MJ), pulsed power facility which is being constructed at Los Alamos National Laboratory with a scheduled completion date in the year 2000. When operational, this facility will provide a platform for experiments in high pressure shocks (>20 Mbar), adiabatic compression ( rho / rho /sub 0/>5, P>10 Mbar), high magnetic fields (~2000 T), high strain and strain rates ( epsilon >200, d epsilon /dt~10/sup 4/ to 10/sup 6/ s/sup -1/), hydrodynamic instabilities of materials in turbulent regimes, magnetized target fusion, equation of state, and strongly coupled plasmas. For the strongly coupled plasma experiments, an auxiliary capacitor bank will be used to generate a moderate density (<0.1 solid), relatively cold (~1 eV) plasma by ohmic heating of a conducting material of interest such as titanium. This target plasma will be compressed against a central column conta...

  1. Experimental research on molten salt thermofluid technology using a high-temperature molten salt loop applied for a fusion reactor Flibe blanket

    Energy Technology Data Exchange (ETDEWEB)

    Toda, Saburo; Chiba, Shinya E-mail: schiba@karma.qse.tohoku.ac.jp; Yuki, Kazuhisa; Omae, Masahiro; Sagara, Akio

    2002-12-01

    Experimental research on molten salt thermofluid technology using a high-temperature molten salt loop (MSL) is described in this paper. The MSL was designed to be able to use Flibe as a coolant, however, a simulant, heat transfer salt (HTS) has to be used alternatively since Flibe is difficult to operate under avoiding a biohazard of Be. Experiment on heat-transfer enhancement, that is required for applying to cool the high heat flux components of fusion reactors, is ongoing. Preliminary experimental results showed that an internal structure of a mixing chamber in the MSL was important to obtain accurate bulk temperatures under severe thermal conditions. For operating the loop, careful handling are needed to proceed how to melt the salt and to circulate it in starting the operation of the MSL. It is concluded that several improvements proposed from the present experiences should be applied for the future Flibe operation.

  2. H Reactor

    Data.gov (United States)

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

  3. Precipitate hydrolysis experimental facility (PHEF) Run 66 And 67, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Shah, H. B.; Lambert, D. P.

    2000-02-11

    analysis. (3) The calculated reactor cooling coil heat transfer coefficient during cooldown was 244 (design basis = 80) PCU/ft{sup 2}/hr/C at 55 C. It took 90 minutes to cooldown the reactor contents at cooling water flowrate of 20 gpm and 168 RPM agitator speed. This suggested that minimal high boilers deposition had occurred with the processing of decomposed feed. (4) Phenylboronic Acid concentration dropped below 53 mg/L in less than 180 minutes. As such very little PBA was produced because of the decomposed feed. The Organic Evaporator (OE) cycle was completed in two attempts. In the first attempt, the OE run was interrupted due to a high Evaporator Condenser Decanter (ECD) vapor temperature interlock, approximately 18 minutes into the aqueous boil (AB). The cause of this interlock was determined to be fouling of the ECD tubes by high boiling organics. Several process changes were implemented in an attempt to defoul the condenser tubes. The changes were mainly targeted at running the condenser hot. The increase in CW supply temperature and reduction in CW flow helped to run the condenser hot, to increase the heat transfer efficiency and to melt the high boiling organics from the condenser tubes.

  4. The National Ignition Facility: Status and Plans for Laser Fusion and High-Energy-Density Experimental Studies

    Energy Technology Data Exchange (ETDEWEB)

    Moses, E I; Wuest, C R

    2002-10-16

    The National Ignition Facility (NIF), currently under construction at the University of California's Lawrence Livermore National Laboratory, is a stadium-sized facility containing a 192-beam, 1.8-Megajoule, 500-Terawatt, 351-nm laser system and a 10-meter diameter target chamber with room for nearly 100 experimental diagnostics. NIF is being built by the National Nuclear Security Administration and when completed will be the world's largest laser experimental system, providing a national center to study inertial confinement fusion and the physics of matter at extreme energy densities and pressures. NIF will provide 192 energetic laser beams that will compress small fusion targets to conditions where they will ignite and burn, liberating more energy than is required to initiate the fusion reactions. NIF experiments will allow the study of physical processes at temperatures approaching 100 million K and 100 billion times atmospheric pressure. These conditions exist naturally only in the interior of stars and in nuclear weapons explosions. In the course of designing the world's most energetic laser system, a number of significant technology breakthroughs have been achieved. Research is also underway to develop a shorter pulse capability on NIF for very high power and extreme electromagnetic field research and applications. We discuss here the technology challenges and solutions that have made NIF possible, along with enhancements to NIF's design that could lead to near-exawatt power levels.

  5. SB-LOCA beyond the design basis in a PWR experimental verification of am procedures in the PKL test facility

    Energy Technology Data Exchange (ETDEWEB)

    Mull, T.; Schoen, B.; Umminger, K.; Wegner, R. [Framatome ANP GmbH, Erlangen (Germany)

    2001-07-01

    The integral test facility PKL at the Technical Center of Framatome ANP (formerly Siemens/KWU) in Erlangen, Germany, simulates a 1300 MWe western type PWR. It is scaled by 1:145 in power and volume at original elevations. It features the entire primary side including four symmetrically arranged coolant loops and auxiliary and safety systems as well as the major part of the secondary side. The test series PKL III D, which was finished at the end of 1999, aimed at the exploration of safety margins and at the efficiency and optimization of operator initiated accident management (AM) procedures. Among others, several tests with small primary breaks combined with additional system failures were performed. This presentation describes test D3.1. The scenario under investigation was a small primary break (24 cm{sup 2} ) with simultaneous failure of the high pressure safety injection (HPSI), a beyond-design-basis scenario. For the German 1300 MWe PWRs, under such additional failure conditions, SB-LOCAs with leak sizes below 25 cm{sup 2} account for 18 % of the integral core damage frequency (CDF). This integral CDF can be estimated to be 3.1*10{sup -6} per year if no credit is taken from AM procedures. The break location in the test under consideration was in the cold leg between reactor coolant pump (RCP) and reactor pressure vessel (RPV). The assumed aggravating circumstances were HPSI failure and unavailability of 2 steam generators (SGs) as well as 3 out of 4 main steam relief and control valves (MS-RCV). The extra borating system was switched to injection mode at low pressurizer level but, by itself, would have been unable to maintain enough coolant to avoid core being uncovered before the pressure reached the setpoint of the accumulators (ACCs). The accident was managed by additional utilization of the chemical- and volume control system (CVCS) to inject water to partly neutralize the leak rate. The plant could be cooled down by 2 SGs using only one MS-RCV. The

  6. SDG and E/ERDA geothermal loop experimental facility. Bi-monthly report, January--February 1977

    Energy Technology Data Exchange (ETDEWEB)

    Bishop, H.K.; Bricarello, J.R.; Enos, F.L.; Hodgdon, N.C.; Jacobson, W.O.; Li, K.K.; Mulliner, D.K.; Swanson, C.R.

    1977-03-01

    Progress is reported in research on the Geothermal Loop Experimental Facility (GLEF), which operated for 758 hours during January and February 1977 with geothermal brine from both Magmamax No. 1 and Woolsey No. 1 wells. The facility was shut down for two weeks for the first major cleaning of the plant. The cleaning was accomplished by hand-scraping the large vessels, hydroblasting (high pressure water jets) portions of the reinjection line, and acid soaking, followed in most instances by hydroblasting, of the facility brine drain lines. The deepening of Woolsey No. 1 well to approximately 3490 feet was completed. Flow tests of Woolsey were attempted, with resumption of facility operation utilizing Magmamax No. 3 as the reinjection well. Insufficient geothermal fluid pressure at the Woolsey No. 1 wellhead resulted in several intermittent operations. Remedial measures undertaken included high pressure injection (hydrofracture) using a separate pump on Magmamax No. 2 and back flushing using a nitrogen purge on Magmamax No. 3. After these remedial measures were taken, the injection-well back pressure remained high, thus limiting the reinjection operation. Downhole data under these low flow conditions were gathered in order to determine further remedial actions. Typical reinjection conditions during this time were 400 gpm, 450 psig and 200/sup 0/F at the plant boundary. Final Woolsey No. 1 wellhead conditions were approximately 200 psig and 400/sup 0/F, indicating that an adequate supply of geothermal brine to the GLEF is very probable. No change in the GLEF brine and steam chemistry occurred during this reporting period. Scaling occurred throughout the system with the greatest deposition in the injection pump and lines. Scale composition varied from an iron, lead, silica mixture in the initial plant stages to a predominantly silica scale in the last stage and injection line. Installation of the electronic data logging system was completed. (JGB)

  7. Experimental results from magnetized-jet experiments executed at the Jupiter Laser Facility

    Science.gov (United States)

    Manuel, M. J.-E.; Kuranz, C. C.; Rasmus, A. M.; Klein, S. R.; MacDonald, M. J.; Trantham, M. R.; Fein, J. R.; Belancourt, P. X.; Young, R. P.; Keiter, P. A.; Drake, R. P.; Pollock, B. B.; Park, J.; Hazi, A. U.; Williams, G. J.; Chen, H.

    2015-12-01

    Recent experiments at the Jupiter Laser Facility investigated magnetization effects on collimated plasma jets. Laser-irradiated plastic-cone-targets produced collimated, millimeter-scale plasma flows as indicated by optical interferometry. Proton radiography of these jets showed no indication of strong, self-generated magnetic fields, suggesting a dominantly hydrodynamic collimating mechanism. Targets were placed in a custom-designed solenoid capable of generating field strengths up to 5 T. Proton radiographs of the well-characterized B-field, without a plasma jet, suggested an external source of trapped electrons that affects proton trajectories. The background magnetic field was aligned with the jet propagation direction, as is the case in many astrophysical systems. Optical interferometry showed that magnetization of the plasma results in disruption of the collimated flow and instead produces a hollow cavity. This result is a topic of ongoing investigation.

  8. Summaries of FY16 LANL experimental campaigns at the OMEGA and EP Laser Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Loomis, Eric Nicholas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Merritt, Elizabeth Catherine [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Montgomery, David [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kim, Yong Ho [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Murphy, Thomas Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Johns, Heather Marie [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kline, John L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Shah, Rahul C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zylstra, Alex [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Herrmann, Hans W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schmitt, Mark J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Flippo, Kirk Adler [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rasmus, Alexander Martin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-10-25

    In FY16, Los Alamos National Laboratory carried out 22 shot days on the OMEGA and OMEGA- EP laser facilities in the areas of High Energy Density (HED) Science and Inertial Confinement Fusion (ICF). In HED our focus areas were on radiation flow, hydrodynamic turbulent mix and burn, warm dense matter equations of state, and coupled Kelvin-­Helmholtz (KH)/Richtmyer-­ Meshkov (RM) instability growth. For ICF our campaigns focused on the Priority Research Directions (PRD) of implosion phase mix and stagnation and burn, specifically as they pertain to Laser Direct Drive (LDD). We also had several focused shot days on transport properties in the kinetic regime. We continue to develop advanced diagnostics such as Neutron Imaging, Gamma Reaction History, and Gas Cherenkov Detectors. Below are a summary of our campaigns, their motivation, and main results from this year.

  9. Design and construction of the ATEFA facility for experimental investigations of AMTEC test modules

    Science.gov (United States)

    Diez de los Rios Ramos, N.; Hering, W.; Weisenburger, A.; Stüber, M.; Onea, A.; Lux, M.; Ulrich, S.; Stieglitz, R.

    2017-07-01

    The Alkali Metal Thermal-to-Electric Converter (AMTEC) is an electrochemical cell that requires a high temperature heat source to generate electricity. At KIT the AMTEC technology is being investigated focusing on the use of concentrating solar energy as heat source. First a review on AMTEC technology is given. Further, the design and realization phases of the AMTEC Test Facility (ATEFA) and AMTEC test cell are presented, including the data acquisition and control system and two key technology developments: a ceramic to metal joint for high temperatures (800 - 1000 °C) and the magnetron sputtering of cathode layers on the ceramic electrolyte. The sheet resistance of several electrode samples has been analyzed using the 4-point probe technique and the microstructure of the cathode layer has been examined using the scanning electron microscopy (SEM).

  10. Experimental neutron capture data of 58Ni from the CERN n_TOF facility

    Directory of Open Access Journals (Sweden)

    Žugec P.

    2015-01-01

    Full Text Available The neutron capture cross section of 58Ni was measured at the neutron time of flight facility n_TOF at CERN, from 27 meV to 400 keV neutron energy. Special care has been taken to identify all the possible sources of background, with the so-called neutron background obtained for the first time using high-precision GEANT4 simulations. The energy range up to 122 keV was treated as the resolved resonance region, where 51 resonances were identified and analyzed by a multilevel R-matrix code SAMMY. Above 122 keV the code SESH was used in analyzing the unresolved resonance region of the capture yield. Maxwellian averaged cross sections were calculated in the temperature range of kT = 5 – 100 keV, and their astrophysical implications were investigated.

  11. Experimental neutron capture data of 58Ni from the CERN n_TOF facility

    Science.gov (United States)

    Žugec, P.; Barbagallo, M.; Colonna, N.; Bosnar, D.; Altstadt, S.; Andrzejewski, J.; Audouin, L.; Bécares, V.; Bečvář, F.; Belloni, F.; Berthoumieux, E.; Billowes, J.; Boccone, V.; Brugger, M.; Calviani, M.; Calviño, F.; Cano-Ott, D.; Carrapiço, C.; Cerutti, F.; Chiaveri, E.; Chin, M.; Cortés, G.; Cortés-Giraldo, M. A.; Diakaki, M.; Domingo-Pardo, C.; Dressler, R.; Duran, I.; Dzysiuk, N.; Eleftheriadis, C.; Ferrari, A.; Fraval, K.; Ganesan, S.; García, A. R.; Giubrone, G.; Gómez-Hornillos, M. B.; Gonçalves, I. F.; González-Romero, E.; Griesmayer, E.; Guerrero, C.; Gunsing, F.; Gurusamy, P.; Heinitz, S.; Jenkins, D. G.; Jericha, E.; Kadi, Y.; Käppeler, F.; Karadimos, D.; Kivel, N.; Koehler, P.; Kokkoris, M.; Krtička, M.; Kroll, J.; Langer, C.; Lederer, C.; Leeb, H.; Leong, L. S.; Lo Meo, S.; Losito, R.; Manousos, A.; Marganiec, J.; Martínez, T.; Massimi, C.; Mastinu, P. F.; Mastromarco, M.; Meaze, M.; Mendoza, E.; Mengoni, A.; Milazzo, P. M.; Mingrone, F.; Mirea, M.; Mondalaers, W.; Paradela, C.; Pavlik, A.; Perkowski, J.; Pignatari, M.; Plompen, A.; Praena, J.; Quesada, J. M.; Rauscher, T.; Reifarth, R.; Riegov, A.; Roman, F.; Rubbia, C.; Sarmento, R.; Saxena, A.; Schillebeeckx, P.; Schmidt, S.; Schumann, D.; Tagliente, G.; Tain, J. L.; Tarrío, D.; Tassan-Got, L.; Tsinganis, A.; Valenta, S.; Vannini, G.; Variale, V.; Vaz, P.; Ventura, A.; Versaci, R.; Vermeulen, M. J.; Vlachoudis, V.; Vlastou, R.; Wallner, A.; Ware, T.; Weigand, M.; Weiß, C.; Wright, T.

    2015-05-01

    The neutron capture cross section of 58Ni was measured at the neutron time of flight facility n_TOF at CERN, from 27 meV to 400 keV neutron energy. Special care has been taken to identify all the possible sources of background, with the so-called neutron background obtained for the first time using high-precision GEANT4 simulations. The energy range up to 122 keV was treated as the resolved resonance region, where 51 resonances were identified and analyzed by a multilevel R-matrix code SAMMY. Above 122 keV the code SESH was used in analyzing the unresolved resonance region of the capture yield. Maxwellian averaged cross sections were calculated in the temperature range of kT = 5 - 100 keV, and their astrophysical implications were investigated.

  12. Cold Climate Foundation Retrofit Experimental Hygrothermal Performance. Cloquet Residential Research Facility Laboratory Results

    Energy Technology Data Exchange (ETDEWEB)

    Goldberg, Louise F. [NorthernSTAR, St. Paul, MN (United States); Harmon, Anna C. [NorthernSTAR, St. Paul, MN (United States)

    2015-04-09

    This project was funded jointly by the National Renewable Energy Laboratory (NREL) and Oak Ridge National Laboratory (ORNL). ORNL focused on developing a full basement wall system experimental database to enable others to validate hygrothermal simulation codes. NREL focused on testing the moisture durability of practical basement wall interior insulation retrofit solutions for cold climates. The project has produced a physically credible and reliable long-term hygrothermal performance database for retrofit foundation wall insulation systems in zone 6 and 7 climates that are fully compliant with the performance criteria in the 2009 Minnesota Energy Code. These data currently span the period from November 10, 2012 through May 31, 2014 and are anticipated to be extended through November 2014. The experimental data were configured into a standard format that can be published online and that is compatible with standard commercially available spreadsheet and database software.

  13. Development of Flow-Through Polymeric Membrane Reactor for Liquid Phase Reactions: Experimental Investigation and Mathematical Modeling

    Directory of Open Access Journals (Sweden)

    Endalkachew Chanie Mengistie

    2017-01-01

    Full Text Available Incorporating metal nanoparticles into polymer membranes can endow the membranes with additional functions. This work explores the development of catalytic polymer membrane through synthesis of palladium nanoparticles based on the approaches of intermatrix synthesis (IMS inside surface functionalized polyethersulfone (PES membrane and its application to liquid phase reactions. Flat sheet PES membranes have been successfully modified via UV-induced graft polymerization of acrylic acid monomer. Palladium nanoparticles have been synthesized by chemical reduction of palladium precursor loaded on surface modified membranes, an approach to the design of membranes modified with nanomaterials. The catalytic performances of the nanoparticle incorporated membranes have been evaluated by the liquid phase reduction of p-nitrophenol using NaBH4 as a reductant in flow-through membrane reactor configuration. The nanocomposite membranes containing palladium nanoparticles were catalytically efficient in achieving a nearly 100% conversion and the conversion was found to be dependent on the flux, amount of catalyst, and initial concentration of nitrophenol. The proposed mathematical model equation represents satisfactorily the reaction and transport phenomena in flow-through catalytic membrane reactor.

  14. Hot Fuel Examination Facility/South

    Energy Technology Data Exchange (ETDEWEB)

    1990-05-01

    This document describes the potential environmental impacts associated with proposed modifications to the Hot Fuel Examination Facility/South (HFEF/S). The proposed action, to modify the existing HFEF/S at the Argonne National Laboratory-West (ANL-W) on the Idaho National Engineering Laboratory (INEL) in southeastern Idaho, would allow important aspects of the Integral Fast Reactor (IFR) concept, offering potential advantages in nuclear safety and economics, to be demonstrated. It would support fuel cycle experiments and would supply fresh fuel to the Experimental Breeder Reactor-II (EBR-II) at the INEL. 35 refs., 12 figs., 13 tabs.

  15. Development of the radiation models of a BWR type reactor and it facility in the SUN-RAH; Desarrollo de modelos de radiacion de un reactor tipo BWR y su instalacion en el SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Barron A, I. [Facultad de Ingenieria, UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: isbarron@yahoo.com.mx

    2005-07-01

    This work about generation models, transport in processes and radioactive contamination of areas of a BWR central, is an amplification to the project developed in the UNAM to have a support tool in subjects or electric generation courses. It is planned about the implementation of models of radiation generation in a BWR type reactor for complement the functions developed in the University Simulator of Nucleo electric- Boiling water reactor (SUN-RAH) which it has been implemented in Simulink of MatLab and it has a model for the dynamics of one nucleo electric central that presents the main characteristics of the reactor vessel, the recirculation system, steam lines, turbines, generator, condensers and feeding water, defined by the main processes that intervene in the generation of energy of these plants. By this way the radiation monitoring systems for area and process, operate simultaneously with the processes of energy generation, with that is possible to observe the changes that present with respect to the operation conditions of the plant, and likewise to appreciate the radiation transport process through the components of the reactor, steam lines and turbines, for different operation conditions and possible faults that they could be presented during the reactor operation. (Author)

  16. Determination of the optimal positions for installing gamma ray detection systems at Tehran Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sayyah, A. [Department of Radiation Application, Shahid Beheshti University (Iran, Islamic Republic of); Rahmani, F., E-mail: FRahamni@kntu.ac.in [K.N. Toosi University of Technology, Department of Physics (Iran, Islamic Republic of); Khalafi, H. [Nuclear Science and Technology Research Institute (NSTRI) (Iran, Islamic Republic of)

    2015-09-01

    Dosimetric instruments must constantly monitor radiation dose levels in different areas of nuclear reactor. Tehran Research Reactor (TRR) has seven beam tubes for different research purposes. All the beam tubes extend from the reactor core to Beam Port Floor (BPF) of the reactor facility. During the reactor operation, the gamma rays exiting from each beam tube outlet produce a specific gamma dose rate field in the space of the BPF. To effectively monitor the gamma dose rates on the BPF, gamma ray detection systems must be installed in optimal positions. The selection of optimal positions is a compromise between two requirements. First, the installation positions must possess largest gamma dose rates and second, gamma ray detectors must not be saturated in these positions. In this study, calculations and experimental measurements have been carried out to identify the optimal positions of the gamma ray detection systems. Eight three dimensional models of the reactor core and related facilities corresponding to eight scenarios have been simulated using MCNPX Monte Carlo code to calculate the gamma dose equivalent rate field in the space of the BPF. These facilities are beam tubes, thermal column, pool, BPF space filled with air, facilities such as neutron radiography facility, neutron powder diffraction facility embedded in the beam tubes as well as biological shields inserted into the unused beam tubes. According to the analysis results of the combined gamma dose rate field, three positions on the north side and two positions on the south side of the BPF have been recognized as optimal positions for installing the gamma ray detection systems. To ensure the consistency of the simulation data, experimental measurements were conducted using TLDs (600 and 700) pairs during the reactor operation at 4.5 MW.

  17. Results from phase 2 of the radioiodine test facility experimental program

    Energy Technology Data Exchange (ETDEWEB)

    Ball, J.M.; Kupferschmid, W.C.H.; Wren, J.C. [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-01

    A series of intermediate-scale experiments were conducted in the Radioiodine Test Facility (RTF) in a vinyl-painted, zinc-primer coated, carbon steel vessel in order to assess the effects of vinyl surfaces on iodine volatility in both the presence and absence of radiation. This test series, Phase 2 of a larger, comprehensive program assessing a variety of containment surfaces, also examined the effects of organic (i.e., methyl ethyl ketone) and inorganic (i.e., hydrazine) additives, pH, and venting on the aqueous chemistry and volatility of solutions initially containing cesium iodide. These tests have clearly demonstrated that organics are released to the aqueous phase from the vinyl coating and that, under radiation conditions, these organics can have a significant effect on the formation of volatile iodine species. In particular, the RTF results suggest that radiolytic decomposition of the released organics results in dramatic reductions in pH and dissolved oxygen concentration, which in turn are responsible for increased formation of molecular iodine and organic iodides. When the pH was maintained at 10 (Test 3), much lower iodine volatility was observed; low iodine volatility was also observed in the absence of radiation. This test series also demonstrated that vinyl surfaces, particularly those in contact with the gas phase, were a sink for iodine. (author) 4 figs., 6 tabs., 17 refs.

  18. Status and Plans for the National Spherical Torus Experimental Research Facility

    Energy Technology Data Exchange (ETDEWEB)

    M. Ono; M.G. Bell; R.E. Bell; J.M. Bialek; T. Bigelow; M. Bitter; plus 148 additional authors

    2005-07-27

    An overview of the research capabilities and the future plans on the MA-class National Spherical Torus Experiment (NSTX) at Princeton is presented. NSTX research is exploring the scientific benefits of modifying the field line structure from that in more conventional aspect ratio devices, such as the tokamak. The relevant scientific issues pursued on NSTX include energy confinement, MHD stability at high beta, non-inductive sustainment, solenoid-free start-up, and power and particle handling. In support of the NSTX research goal, research tools are being developed by the NSTX team. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a high beta Demo device based on the ST, are being considered. For these, it is essential to develop high performance (high beta and high confinement), steady-state (non-inductively driven) ST operational scenarios and an efficient solenoid-free start-up concept. We will also briefly describe the Next-Step-ST (NSST) device being designed to address these issues in fusion-relevant plasma conditions.

  19. Experimental observations of turbulent mixing due to Kelvin-Helmholtz instability on the OMEGA Laser Facility

    Energy Technology Data Exchange (ETDEWEB)

    Smalyuk, V. A.; Hansen, J. F.; Hurricane, O. A.; Langstaff, G.; Martinez, D.; Park, H.-S.; Raman, K.; Remington, B. A.; Robey, H. F.; Schilling, O.; Wallace, R. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States); Elbaz, Y.; Shimony, A.; Shvarts, D. [Physics Department, Nuclear Research Center Negev, Beer-Sheva (Israel); Physics Department, Ben-Gurion University, Beer-Sheva (Israel); Di Stefano, C.; Drake, R. P.; Marion, D.; Krauland, C. M.; Kuranz, C. C. [Department of Atmospheric, Oceanic, and Space Sciences, University of Michigan, Ann Arbor, Michigan 48109 (United States)

    2012-09-15

    Shear-flow, Kelvin-Helmholtz (KH) turbulent mixing experiments were performed on the OMEGA Laser Facility [Boehly et al., Opt. Commun. 133, 495 (1997)] in which laser-driven shock waves propagated through a low-density plastic foam placed on top of a higher-density plastic foil. The plastic foil was comprised a thin iodine-doped plastic tracer layer bonded on each side to an undoped density-matched polyamide-imide plastic. Behind the shock front, lower-density foam plasma flowed over the higher-density plastic plasma, such that the interface between the foam and plastic was KH unstable. The initial perturbations consisted of pre-imposed, sinusoidal 2D perturbations, and broadband 3D perturbations due to surface roughness at the interface between the plastic and foam. KH instability growth was measured using side-on radiography with a point-projection 5-keV vanadium backlighter. Time-integrated images were captured on D-8 x-ray film. Spatial density profiles of iodine-doped plastic mixed with foam were inferred using x-ray radiographs. The mixing layer ensuing from the KH instability with layer width up to {approx}100 {mu}m was observed at a location {approx}1 mm behind the shock front. The measured mixing layer width was in good agreement with predictions based on a simple self-similar model of KH instability growth using an estimate of the shear velocity obtained from numerical simulations of the experiments.

  20. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor - FY-05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2005-09-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 9000C or operational fuel temperatures above 12500C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR’s higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Laboratory (INL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world’s computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertainty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  1. Experimental infections with rifampicin-resistant Clostridium perfringens strains in broiler chickens using isolator facilities

    DEFF Research Database (Denmark)

    Pedersen, Karl; Bjerrum, Lotte; Nauerby, Birgitte

    2003-01-01

    Experimental infection studies were carried out on the ability of three Clostridium perfringens type A rifampicin-resistant strains to colonize the intestinal tract of broiler chickens kept in isolators from 1-day-old. Various doses of C. perfringens were given orally at 22 days, 9 days or at 1 day...... replaced by naturally occurring strains of C. perfringens in all groups but they persisted for considerably longer in chickens inoculated at 1-day-old or at 9 days than those at 22 days, indicating a possible resistance to colonization with increasing age. The findings emphasize the difficulties...... of establishing a reproducible model for infection with C. perfringens in broiler chickens....

  2. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Directory of Open Access Journals (Sweden)

    Hwang Bae

    2017-08-01

    Full Text Available Three small-break loss-of-coolant accident (SBLOCA tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor, i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  3. Comparison of three small-break loss-of-coolant accident tests with different break locations using the system-integrated modular advanced reactor-integral test loop facility to estimate the safety of the smart design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Hwang; Ryu, Sung Uk; Yi, Sung Jae; Park, Hyun Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Dept. of Precision Mechanical Engineering, Kyungpook National University, Sangju (Korea, Republic of)

    2017-08-15

    Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal–hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

  4. Online optimal experimental re-design in robotic parallel fed-batch cultivation facilities.

    Science.gov (United States)

    Cruz Bournazou, M N; Barz, T; Nickel, D B; Lopez Cárdenas, D C; Glauche, F; Knepper, A; Neubauer, P

    2017-03-01

    We present an integrated framework for the online optimal experimental re-design applied to parallel nonlinear dynamic processes that aims to precisely estimate the parameter set of macro kinetic growth models with minimal experimental effort. This provides a systematic solution for rapid validation of a specific model to new strains, mutants, or products. In biosciences, this is especially important as model identification is a long and laborious process which is continuing to limit the use of mathematical modeling in this field. The strength of this approach is demonstrated by fitting a macro-kinetic differential equation model for Escherichia coli fed-batch processes after 6 h of cultivation. The system includes two fully-automated liquid handling robots; one containing eight mini-bioreactors and another used for automated at-line analyses, which allows for the immediate use of the available data in the modeling environment. As a result, the experiment can be continually re-designed while the cultivations are running using the information generated by periodical parameter estimations. The advantages of an online re-computation of the optimal experiment are proven by a 50-fold lower average coefficient of variation on the parameter estimates compared to the sequential method (4.83% instead of 235.86%). The success obtained in such a complex system is a further step towards a more efficient computer aided bioprocess development. Biotechnol. Bioeng. 2017;114: 610-619. © 2016 Wiley Periodicals, Inc. © 2016 Wiley Periodicals, Inc.

  5. Cold Climate Foundation Retrofit Experimental Hygrothermal Performance: Cloquet Residential Research Facility Laboratory Results

    Energy Technology Data Exchange (ETDEWEB)

    Goldberg, Louise F. [NorthernSTAR, St. Paul, MN (United States); Harmon, Anna C. [NorthernSTAR, St. Paul, MN (United States)

    2015-04-01

    Thermal and moisture problems in existing basements create a unique challenge because the exterior face of the wall is not easily or inexpensively accessible. This approach addresses thermal and moisture management from the interior face of the wall without disturbing the exterior soil and landscaping. the interior and exterior environments. This approach has the potential for improving durability, comfort, and indoor air quality. This project was funded jointly by the National Renewable Energy Laboratory (NREL) and Oak Ridge National Laboratory (ORNL). ORNL focused on developing a full basement wall system experimental database to enable others to validate hygrothermal simulation codes. NREL focused on testing the moisture durability of practical basement wall interior insulation retrofit solutions for cold climates. The project has produced a physically credible and reliable long-term hygrothermal performance database for retrofit foundation wall insulation systems in zone 6 and 7 climates that are fully compliant with the performance criteria in the 2009 Minnesota Energy Code. The experimental data were configured into a standard format that can be published online and that is compatible with standard commercially available spreadsheet and database software.

  6. Experimental and modeling study of the effect of CO and H2 on the urea DeNO(x) process in a 150kW laboratory reactor.

    Science.gov (United States)

    Javed, M Tayyeb; Nimmo, W; Gibbs, B M

    2008-01-01

    An experimental and modeling investigation has been performed to study the effect of process additives, H2 and CO on NO(x) removal from flue gases by a selective non-catalytic reduction process using urea as a reducing agent. Experiments were performed with a flow reactor in which flue gas was generated by the combustion of propane in air at 3% excess oxygen and the desired levels of initial NO(x) (500ppm) were achieved by doping the flame with ammonia. Experiments were performed throughout the temperature range of interest, i.e. from 850 to 1200 degrees C for investigation of the effects of the process additives on the performance of aqueous urea DeNO(x). Subsequently, computational kinetic modeling with SENKIN code was performed to analyze the performance of urea providing a direct comparison of modeling prediction with experimental measurements. With CO addition, a downwards shift of 215 degrees C in the peak reduction temperature from 1125 to 910 degrees C was observed during the experimentation while the kinetic modeling suggests it to be 150 degrees C, i.e. from 1020 to 870 degrees C. The addition of H2 impairs the peak NO(x) reduction but suggests a low temperature application of the process. A downward shift of 250 degrees C in the peak reduction temperature, from 1020 to 770 degrees C, was observed during kinetic modeling studies. The kinetic modeling shows a good qualitative agreement with the experimental observations and reveals additional information about the process.

  7. Sequential Aeration of Membrane-Aerated Biofilm Reactors for High-Rate Autotrophic Nitrogen Removal: Experimental Demonstration

    DEFF Research Database (Denmark)

    Pellicer i Nàcher, Carles; Sun, Sheng-Peng; Lackner, Susanne

    2010-01-01

    One-stage autotrophic nitrogen (N) removal, requiring the simultaneous activity of aerobic and anaerobic ammonium oxidizing bacteria (AOB and AnAOB), can be obtained in spatially redox-stratified biofilms. However, previous experience with Membrane-Aerated Biofilm