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Sample records for gas plenum experiment

  1. Requirements for the GCFR plenum streaming experiment

    International Nuclear Information System (INIS)

    Perkins, R.G.; Rouse, C.A.; Hamilton, C.J.

    1980-09-01

    This report gives the experiment objectives and generic descriptions of experimental configurations for the gas-cooled fast breeder reactor (GCFR) plenum shield experiment. This report defines four experiment phases. Each phase represents a distinct area of uncertainty in computing radiation transport from the GCFR core to the plenums, through the upper and lower plenum shields, and ultimately to the prestressed concrete reactor vessel (PCRV) liner: (1) the shield heterogeneity phase; (2) the exit shield simulation phase; (3) the plenum streaming phase; and (4) the plenum shield simulation phase

  2. Scaled Experimental Modeling of VHTR Plenum Flows

    Energy Technology Data Exchange (ETDEWEB)

    ICONE 15

    2007-04-01

    Abstract The Very High Temperature Reactor (VHTR) is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S. which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. Various scaled heated gas and water flow facilities were investigated for modeling VHTR upper and lower plenum flows during the decay heat portion of a pressurized conduction-cooldown scenario and for modeling thermal mixing and stratification (“thermal striping”) in the lower plenum during normal operation. It was concluded, based on phenomena scaling and instrumentation and other practical considerations, that a heated water flow scale model facility is preferable to a heated gas flow facility and to unheated facilities which use fluids with ranges of density to simulate the density effect of heating. For a heated water flow lower plenum model, both the Richardson numbers and Reynolds numbers may be approximately matched for conduction-cooldown natural circulation conditions. Thermal mixing during normal operation may be simulated but at lower, but still fully turbulent, Reynolds numbers than in the prototype. Natural circulation flows in the upper plenum may also be simulated in a separate heated water flow facility that uses the same plumbing as the lower plenum model. However, Reynolds number scaling distortions will occur at matching Richardson numbers due primarily to the necessity of using a reduced number of channels connected to the plenum than in the prototype (which has approximately 11,000 core channels connected to the upper plenum) in an otherwise geometrically scaled model. Experiments conducted in either or both facilities will meet the objectives of providing benchmark data for the validation of codes proposed for NGNP designs and safety studies, as well as providing a better understanding of the complex flow phenomena in the plenums.

  3. Experiments on the lower plenum response during a severe accident

    International Nuclear Information System (INIS)

    Henry, Robert E.; Hammersley, Robert J.; Klopp, George T.; Merilo, Mati

    2004-01-01

    Severe accident evaluations for nuclear reactors consider the response when the core materials have been overheated sufficient to melt and change geometry. One possible consequence of this is that molten core debris could drain into the lower plenum, as occurred in the TMI-2 accident. Given this state, several physical processes need to be analyzed, i.e. the extent of debris particulation and cooling, the potential for thermal attack of lower plenum structures, the thermal transient of the RPV and the potential for external cooling of the RPV lower head. These are important and complex processes, the evaluations of which need to be guided by well founded experiments. To support the development of the MAAP codes, recent experiments have been performed on specific issues such as: 1. the response of lower head penetrations submerged in a high temperature melt, 2. the net steam generation rate when molten debris drains into the lower plenum, 3. the formation of a contact resistance when molten debris drains through water and contacts the RPV wall and 4. the potential for external cooling of the RPV lower head. This paper discusses these experiments and their results. More importantly, it discusses how these are used in formulating models to represent the lower plenum response in the MAAP codes. (author)

  4. Numerical simulation and geometry optimization of hot-gas mixing in lower plenum of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Hang; Wang Jie; Laurien, E.

    2010-01-01

    The lower plenum in high temperature gas-cooled reactor was designed to mix the gas of different temperatures from the reactor core. Previous researches suggest the current geometry of the lower plenum to be improved for better mixing capability and lower pressure drop. In the presented work, a series of varied geometries were investigated with numerical simulation way. The choice of appropriate mesh type and size used in the geometry variation was discussed with the reference of experimental data. The original thin ribs in the current design were merged into thicker ones, and a junction located at the starting end of the outlet pipe was introduced. After comparing several potential optimization methods, an improved geometry was selected with the merged ribs increasing the pre-defined mixing coefficient and the junction reducing the pressure drop. Future work was discussed based on the simulation of real reactor case. The work shows a direction for design improvements of the lower plenum geometry. (authors)

  5. Experimental optimization of temperature distribution in the hot-gas duct through the installation of internals in the hot-gas plenum of a high-temperature reactor

    International Nuclear Information System (INIS)

    Henssen, J.; Mauersberger, R.

    1990-01-01

    The flow conditions in the hot-gas plenum and in the adjacent hot-gas ducts and hot-gas pipes for the high-temperature reactor project PNP-1000 (nuclear process heat project for 1000 MW thermal output) have been examined experimentally. The experiments were performed in a closed loop in which the flow model to be analyzed, representing a 60deg sector of the core bottom of the PNP-1000 with connecting hot-gas piping and diverting arrangements, was installed. The model scale was approx. 1:5.6. The temperature and flow velocity distribution in the hot-gas duct was registered by means of 14 dual hot-wire flowmeters. Through structural changes and/or the installation of internals into the hot-gas plenum of the core bottom offering little flow resistance coolant gas temperature differentials produced in the core could be reduced to such an extent that a degree of mixture amounting to over 80% was achieved at the entrance of the connected heat exchanger systems. Thereby the desired goal of an adequate degree of mixture of the hot gas involving an acceptable pressure loss was reached. (orig.)

  6. Optimization of inlet plenum of A PBMR using surrogate modeling

    International Nuclear Information System (INIS)

    Lee, Sang-Moon; Kim, Kwang-Yong

    2009-01-01

    The purpose of present work is to optimize the design of inlet plenum of PBMR type gas cooled nuclear reactor numerically using a combining of three-dimensional Reynolds-averaged Navier-Stokes (RANS) analysis and surrogate modeling technique. Shear stress transport (SST) turbulence model is used as a turbulence closure. Three geometric design variables are selected, namely, rising channel diameter to plenum height ratio, aspect ratio of the plenum cross section, and inlet port angle. The objective function is defined as a linear combination of uniformity of three-dimensional flow distribution term and pressure drop in the inlet plenum and rising channels of PBMR term with a weighting factor. Twenty design points are selected using Latin-hypercube method of design of experiment and objective function values are obtained at each design point using RANS solver. (author)

  7. Final report on 3-D experiment project air-water upper plenum experiments

    International Nuclear Information System (INIS)

    Jacoby, J.K.; Mohr, C.M.

    1978-11-01

    The results are presented from upper plenum air-water reflood behavior testing performed as part of the program to investigate three-dimensional aspects of PWR LOCA research. Tests described were performed at near ambient temperature and pressure in a plexiglass vessel which included the important features of the upper core and upper plenum regions corresponding to a single fuel bundle in both Westinghouse Electric Corporation (Trojan) and Kraftwerk Union (KKU) PWR designs. The data included observed two-phase flow characteristics, particularly with regard to countercurrent flow, and cinematography of the characteristic upper plenum flow patterns

  8. Design of pellet surface grooves for fission gas plenum

    International Nuclear Information System (INIS)

    Carter, T.J.; Jones, L.R.; Macici, N.; Miller, G.C.

    1986-01-01

    In the Canada deuterium uranium pressurized heavy water reactor, short (50-cm) Zircaloy-4 clad bundles are fueled on-power. Although internal void volume within the fuel rods is adequate for the present once-through natural uranium cycle, the authors have investigated methods for increasing the internal gas storage volume needed in high-power, high-burnup, experimental ceramic fuels. This present work sought to prove the methodology for design of gas storage volume within the fuel pellets - specifically the use of grooves pressed or machined into the relatively cool pellet/cladding interface. Preanalysis and design of pellet groove shape and volume was accomplished using the TRUMP heat transfer code. Postirradiation examination (PIE) was used to check the initial design and heat transfer assumptions. Fission gas release was found to be higher for the grooved pellet rods than for the comparison rods with hollow or unmodified pellets. This had been expected from the initial TRUMP thermal analyses. The ELESIM fuel modeling code was used to check in-reactor performance, but some modifications were necessary to accommodate the loss of heat transfer surface to the grooves. It was concluded that for plenum design purposes, circumferential pellet grooves could be adequately modeled by the codes TRUMP and ELESIM

  9. Plenum separator system for pool-type nuclear reactors

    International Nuclear Information System (INIS)

    Sharbaugh, J.E.

    1983-01-01

    This invention provides a plenum separator system for pool-type nuclear reactors which substantially lessens undesirable thermal effects on major components. A primary feature of the invention is the addition of one or more intermediate plena, containing substantially stagnant and stratified coolant, which separate the hot and cold plena and particularly the hot plena from critical reactor components. This plenum separator system also includes a plurality of components which together form a dual pass flow path annular region spaced from the reactor vessel wall by an annular gas space. The bypass flow through the flow path is relatively small and is drawn from the main coolant pumps and discharged to an intermediate plenum

  10. Development Of An Experiment For Measuring Flow Phenomena Occurring In A Lower Plenum For VHTR CFD Assessment

    Energy Technology Data Exchange (ETDEWEB)

    D. M. McEligot; K.G. Condie; G. E. Mc Creery; H. M. Mc Ilroy

    2005-09-01

    The objective of the present report is to document the design of our first experiment to measure generic flow phenomena expected to occur in the lower plenum of a typical prismatic VHTR (Very High Temperature Reactor) concept. In the process, fabrication sketches are provided for the use of CFD (computational fluid dynamics) analysts wishing to employ the data for assessment of their proposed codes. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/systems code calculations for the same geometry. One aspect of the complex flow in a prismatic VHTR is being addressed: flow and thermal mixing in the lower plenum ("hot streaking" issue). Current prismatic VHTR concepts were examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses were applied to determine key non-dimensional parameters and their magnitudes over this operating range. The flow in the lower plenum can locally be considered to be a situation of multiple jets into a confined crossflow -- with obstructions. Flow is expected to be turbulent with momentum-dominated turbulent jets entering; buoyancy influences are estimated to be negligible in normal full power operation. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments available are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls - with near stagnant surroundings at one extreme and significant crossflow at the other.

  11. Upper plenum dump during reflood in PWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Sudo, Yukio; Griffith, Peter.

    1981-01-01

    Upper plenum dump during reflood in a large break loww-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood. The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnetude of water subcooling. (author)

  12. Influence of graphite discs, chamfers, and plenums on temperature distributions in high burnup fuel

    International Nuclear Information System (INIS)

    Ranger, A.; Tayal, M.; Singh, P.

    1990-04-01

    Previous studies have demonstrated the desirability to increase the fuel burnups in CANDU reactors from 7-9 GW.d/t to 21 GW.d/t. At high burnups, one consideration in fuel integrity is fission gas pressure, which is predicted to reach about 13 MPa. The gas pressure can be kept below the coolant pressure (about 10 MPa) via a variety of options such as bigger chamfers, deeper dishes, central hole, and plenums. However, it is important to address the temperature perturbations produced by the bigger chamfers and plenums which in turn, affect the gas pressure. Another consideration in fuel integrity is to reduce the likelihood of fuel failures via environmentally assisted cracking. Insertion of graphite discs between neighbouring pellets will lower the pellet temperatures, hence, lower fission gas release and lower expansion of the pellet. Therefore, it is desired to quantify the effect of graphite discs on pellet temperatures. Thermal analyses of different fuel element geometries: with and without chamfers, graphite discs, and plenums were performed. The results indicate that the two-dimensional distributions of temperatures due to the presence of chamfers, graphite discs, or plenums can have a significant impact on the integrity of high burnup fuel as we have been able to quantify in this paper

  13. Evaluation of axial fission gas transport in power ramping experiments

    International Nuclear Information System (INIS)

    Kinoshita, Motoyasu

    1986-01-01

    The LINUS code calculates advective and diffusional transport of fission gas towards an upper plenum through the pellet-cladding gap. The basic equations were modified for analyzing a multi-component gas mixture in the gap and also for dealing with opening and/or closing of the gap, which induces additional axial gas flow. Analysis of the Petten ramp experiment shows that helium pressurization is effective in suppressing an ascending rate of fission gas concentration. After the maximum concentration is achieved through power ramping, the gas concentration could be described by a steady state analytical solution which does not depend on the filling gas pressure. (author)

  14. IDAHO NATIONAL LABORATORY PROGRAM TO OBTAIN BENCHMARK DATA ON THE FLOW PHENOMENA IN A SCALED MODEL OF A PRISMATIC GAS-COOLED REACTOR LOWER PLENUM FOR THE VALIDATION OF CFD CODES

    International Nuclear Information System (INIS)

    Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink

    2008-01-01

    The experimental program that is being conducted at the Matched Index-of-Refraction (MIR) Flow Facility at Idaho National Laboratory (INL) to obtain benchmark data on measurements of flow phenomena in a scaled model of a typical prismatic gas-cooled (GCR) reactor lower plenum using 3-D Particle Image Velocimetry (PIV) is presented. A detailed description of the model, scaling, the experimental facility, 3-D PIV system, measurement uncertainties and analysis, experimental procedures and samples of the data sets that have been obtained are included. Samples of the data set that are presented include mean-velocity-field and turbulence data in an approximately 1:7 scale model of a region of the lower plenum of a typical prismatic GCR design. This experiment has been selected as the first Standard Problem endorsed by the Generation IV International Forum. Results concentrate on the region of the lower plenum near its far reflector wall (away from the outlet duct). Inlet jet Reynolds numbers (based on the jet diameter and the time-mean average flow rate) are approximately 4,300 and 12,400. The measurements reveal undeveloped, non-uniform flow in the inlet jets and complicated flow patterns in the model lower plenum. Data include three-dimensional vector plots, data displays along the coordinate planes (slices) and charts that describe the component flows at specific regions in the model. Information on inlet flow is also presented

  15. Upper plenum mixing in a BWR

    International Nuclear Information System (INIS)

    Alamgir, M.; Andersen, J.G.M.; Parameswaran, V.

    1984-01-01

    A model for the emergency core cooling injection into the upper plenum of a boiling water reactor has been formulated and implemented into the TRACB02 computer program. The model consists of a spray model and a submerged jet model. The submerged jet model is used when the spray nozzles are covered by a two-phase mixture, and the spray model is used when the nozzles are uncovered. The upper plenum model has been assessed by comparison to an upper plenum mixing test in the Steam Sector Test Facility. It is found that the model accurately predicts the phenomena in the upper plenum of a boiling water reactor

  16. Experiments on graphite block gaps connected with leak flow in bottom-core structure of experimental very high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Kikuchi, Kenji; Futakawa, Masatoshi; Takizuka, Takakazu; Kaburaki, Hideo; Sanokawa, Konomo

    1984-01-01

    In order to minimize the leak flow rate of an experimental VHTR (a multi-purpose very high-temperature gas-cooled reactor), the graphite blocks are tightened to reduce the gap distance between blocks by core restrainers surrounded outside of the fixed reflectors of the bottom-core structure and seal elements are placed in the gaps. By using a 1/2.75-scale model of the bottom-core structure, the experiments on the following items have been carried out: a relationship between core restraint force and block gap, a relationship between core restraint force and inclined angle of the model, leak flow characteristics of seal elements etc. The conclusions derived from the experiments are as follows: (1) Core restraint force is significantly effective for decreasing the gap distance between hot plenum blocks, but ineffective for the gap between hot plenum block and fixed reflector. (2) Graphite seal element reduces the leak flow rate from the top surface of hot plenum block into plenum region to one-third. (author)

  17. Analytical model for transient fluid mixing in upper outlet plenum of an LMFBR

    International Nuclear Information System (INIS)

    Yang, J.W.; Agrawal, A.K.

    1976-01-01

    A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the outlet plenum of an LMFBR. The maximum penetration of core flow is used as the criterion for dividing the sodium region into two mixing zones. The model considers the transient sodium temperature affected by the thermal expansion of sodium, heat transfer with cover gas, heat capacity of different sections of metal and the addition of by-pass flow into the plenum. The results of numerical calculations indicate that effects of flow stratification, chimney height, metal heat capacity and by-pass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas do not play any significant role on sodium temperature

  18. Validation of turbulence models for LMFBR outlet plenum flows

    International Nuclear Information System (INIS)

    Chen, Y.B.; Golay, M.W.

    1977-01-01

    Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds number (Re) values of 33000 and 70000 in a 1/15 - scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different two-equation turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet flow field, importantly also upon the degree of inlet turbulence, and also upon the turbulent momentum exchange model used in the calculations. In the FFTF geometry, the TEACH-T predictions agree well with the experiments. 7 refs

  19. A CFD Study on Inlet Plenum Flow Field of Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Kim, Min Hwan; Lee, Won Jae; Chang, Jong Hwa

    2005-01-01

    High temperature gas cooled reactor, largely divided into two types of PBR (Pebble Bed Reactor) and PMR (Prismatic Modular Reactor), has becomes great interest of researchers in connection with the hydrogen production. KAERI has started a project to develop the gas cooled reactor for the hydrogen production and has been doing in-depth study for selecting the reactor type between PBR and PMR. As a part of the study, PBMR (Pebble Bed Modular Reactor) was selected as a reference PBR reactor for the CFD analysis and the flow field of its inlet plenum was simulated with computational fluid dynamics program CFX5. Due to asymmetrical arrangement of pipes to the inlet plenum, non-uniform flow distribution has been expected to occur, giving rise to non-uniform power distribution at the core. Flow fields of different arrangement of inlet pipes were also investigated, as one of measures to reduce the non-uniformity

  20. Determination of fission gas release of spent nuclear fuel in puncturing test and in leaching experiments under anoxic conditions

    Energy Technology Data Exchange (ETDEWEB)

    González-Robles, E., E-mail: ernesto.gonzalez-robles@kit.edu [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Metz, V. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Wegen, D.H. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Herm, M. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Papaioannou, D. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Bohnert, E. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Gretter, R. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Müller, N. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany); Nasyrow, R.; Weerd, W. de; Wiss, T. [European Commission, Joint Research Centre, Institute for Transuranium Elements (JRC-ITU), P.O. Box 2340, 76125, Karlsruhe (Germany); Kienzler, B. [Karlsruhe Institute of Technology, Institute for Nuclear Waste Disposal (KIT-INE), P.O. Box 3640, D-76021, Karlsruhe (Germany)

    2016-10-15

    During reactor operation the fission gases Kr and Xe are formed within the UO{sub 2} matrix of nuclear fuel. Their quantification is important to evaluate their impact on critical parameters regarding the fuel behaviour during irradiation and (long-term) interim storage, such as internal pressure of the fuel rod and fuel swelling. Moreover the content of Kr and Xe in the plenum of a fuel rod and their content in the UO{sub 2} fuel itself are widely used as indicators for the release properties of {sup 129}I, {sup 137}Cs, and other safety relevant radionuclides with respect to final disposal of spent nuclear fuel. The present study deals with the fission gas release from spent nuclear fuel exposed to simulated groundwater in comparison with the fission gas previously released to the fuel rod plenum during irradiation in reactor. In a unique approach we determined both the Kr and Xe inventories in the plenum by means of a puncturing test and in leaching experiments with a cladded fuel pellet and fuel fragments in bicarbonate water under 3.2 bar H{sub 2} overpressure. The fractional inventory of the fission gases released during irradiation into the plenum was (8.3 ± 0.9) %. The fraction of inventory of fission gases released during the leaching experiments was (17 ± 2) % after 333 days of leaching of the cladded pellet and (25 ± 2) % after 447 days of leaching of the fuel fragments, respectively. The relatively high release of fission gases in the experiment with fuel fragments was caused by the increased accessibility of water to the Kr and Xe occluded in the fuel.

  1. Coolant mixing in the LMFBR outlet plenum

    International Nuclear Information System (INIS)

    Chen, Y.B.; Golay, M.W.

    1977-06-01

    Small scale experiments involving water flows are used to provide mean flow and turbulence field data for LMFBR outlet plenum flows. Measurements are performed at Reynolds Number (Re) values of 33000 and 70000 in a 1/15-scale FFTF geometry and at Re = 35000 in a 3/80-scale CRBR geometry. The experimental behavior is predicted using two different turbulence model computer programs, TEACH-T and VARR-II. It is found that the qualitative nature of the flow field within the plenum depends strongly upon the distribution of the mean inlet velocity field, upon the degree of inlet turbulence, and upon the turbulence momentum exchange model used in the calculations. It is found in the FFTF geometry that the TEACH-T predictions are better than that of VARR-II, and in the CRBR geometry neither code provides a good prediction of the observed behavior. From the sensitivity analysis, it is found that the production and dissipation of turbulence are the dominant terms in the transport equations for turbulent kinetic energy and turbulent energy dissipation rate, and the diffusion terms are relatively small. From the same study a new set of empirical constants for the turbulence model is evolved for the prediction of plenum flows

  2. Improved plenum pressure gradient facemaps for PKL reactors

    International Nuclear Information System (INIS)

    Crowley, D.A.; Hamm, L.L.

    1988-05-01

    This report documents the development of improved plenum pressure gradient facemaps* for PKL Mark 16--31 and Mark 22 reactor charges. These new maps are based on the 1985 L-area AC flow tests. Use of the L-area data base for estimating C-area plenum pressure gradient maps is inappropriate because the nozzle geometry plays a major role in determining the shape of the plenum pressure profile. These plenum pressure gradient facemaps are used in the emergency cooling system (ECS) and in the flow instability (FI) loss of coolant accident (LOCA) limits calculations. For the ECS LOCA limits calculations, the maps are used as input to the FLOWZONE computer code to determine the average flow within a flowzone during normal operating conditions. For the FI LOCA limits calculations, the maps are used as plenum pressure boundary conditions in the FLOWTRAN computer code to determine the maximum pre-incident assembly flow within a flowzone. These maps will also be used for flowzoning and transient protection limits analyses

  3. RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

    Energy Technology Data Exchange (ETDEWEB)

    Bayless, P.D.

    2003-01-17

    The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior.

  4. RELAP/MOD3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

    International Nuclear Information System (INIS)

    Bayless, P.D.

    2003-01-01

    The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior

  5. Stratification in SNR-300 outlet plenum

    International Nuclear Information System (INIS)

    Reinders, R.

    1983-01-01

    In the inner outlet plenum of the SNR-300 under steady state conditions a large toroidal vortex is expected. The main flow passes through the gap between dipplate and shield vessel to the outer annular space. Only 3% of the flow pass the 24 emergency cooling holes, situated in the shield vessel. The sodium leaves the reactor tank through the 3 symmetrically arranged outlet nozzles. For a scram flow rates and temperatures are decreased simultaneously, so it is expected, that stratification occurs in the inner outlet plenum. A measure of stratification effects is the Archimedes Number Ar, which is the relation of buoyancy forces (negative) to kinetic energy. (The Archimedes Number is nearly identical with the Richardson Number). For values Ar>1 stratification can occur. Under the assumption of stratification the code TIRE was developed, which is only applicable for the period of time after some 50 sec after scram. This code serves for long term calculations. As the equations are very simple, it is a very fast code which gives the possibility to calculate transients for some hours real time. This code mainly has to take into account the pressure difference between inner plenum and outlet annulus caused by geodatic pressure. That force is in equilibrium with the pressure drop over the gap and holes in the shield vessel. For more detailed calculations of flow pattern and temperature distribution the code MIX and INKO 2T are applied. MIX was developed and validated at ANL, INKO 2T is a development of INTERATOM. INKO 2T is under validation. Mock up experiments were carried out with water to simulate the transient behavior of the SNR-300 outlet plenum. Calculations obtained by INKO 2T for steady state and the transient are shown for the flow pattern. Results of measurements also prove that stratification begins after about 30 sec. Measurements and detailed calculations show that it is admissible to use the code TIRE for the long term calculations. Calculations for a scram

  6. Analysis of transient thermal response in the outlet plenum of an LMFBR

    International Nuclear Information System (INIS)

    Yang, J.W.

    1976-05-01

    A two-zone mixing model based on the lumped-parameter approach was developed for the analysis of transient thermal response in the upper outlet plenum of an LMFBR. The one-dimensional turbulent jet flow equations were solved to determine the maximum penetration of the core flow. The maximum penetration is used as the criterion for dividing the sodium region into two mixing zones. The lumped-parameter model considers the transient sodium temperature affected by the thermal expansion of sodium, heat transfer with cover gas, heat capacity of different sections of metal and the addition of bypass flow into the plenum. Numerical calculations were performed for two cases. The first case corresponds to a normal scram followed by flow coast-down. The second case represents the double-ended pipe rupture at the inlet of cold leg followed by reactor scram. The results indicate that effects of flow stratification, chimney height, metal heat capacity and bypass flow are important for transient sodium temperature calculation. Thermal expansion of sodium and heat transfer with the cover gas does not play any significant role on sodium temperature. This two-zone mixing model will be a part of the thermohydraulic transient code SSC

  7. Determining Bond Sodium Remaining in Plenum Region of Spent Nuclear Driver Fuel

    International Nuclear Information System (INIS)

    Vaden, D.; Li, S.X.

    2008-01-01

    The Fuel Conditioning Facility (FCF) at the Idaho National Laboratory (INL) treats spent nuclear fuel using an electro-chemical process that separates the uranium from the fission products, sodium thermal bond, and cladding materials (REF 1). Upon immersion into the ER electrolyte, the sodium used to thermally bond the fuel to the clad jacket chemically reacts with the UCl3 in the electrolyte producing NaCl and uranium metal. The uranium in the spent fuel is separated from the cladding and fission products by taking advantage of the electro-chemical potential differences between uranium and the other fuel components. Assuming all the sodium in the thermal bond is converted to NaCl in the ER, the difference between the cumulative bond sodium mass in the fuel elements and the cumulative sodium mass found in the driver ER electrolyte inventory provides an upper mass limit for the sodium that migrated to the upper gas region, or plenum section, of the fuel element during irradiation in the reactor. The plenums are to be processed as metal waste via melting and metal consolidation operations. However, depending on the amount of sodium in the plenums, additional processing may be required to remove the sodium before metal waste processing

  8. RELAP-5/MOD 3.2 Assessment Using an 11% Upper Plenum Break Experiment in the PSB Facility

    Energy Technology Data Exchange (ETDEWEB)

    Paul D. Bayless

    2003-01-01

    The RELAP/MOD3.2 computer code has been assessed using an 11% upper plenum break experiment in the PSB test facility at the Electrogorsk Research and Engineering Center. This work was performed as part of the U.S. Department of Energy's International Nuclear Safety Program, and is part of the effort addressing the capability of the RELAP5/MOD3.2 code to model transients in Soviet-designed reactors. Designated VVER Standard Problem PSBV1, the test addressed several important phenomena related to VVER behavior that the code needs to simulate well. The code was judged to reasonably model the phenomena of two-phase flow natural circulation in the primary coolant system, asymmetric loop behavior, leak flow, loop seal clearance in the cold legs, heat transfer in a covered core, heat transfer in a partially covered core, pressurizer thermal-hydraulics, and integral system effects. The code was judged to be in minimal agreement with the experiment data for the mixture level and entrainment in the core, leading to a user recommendation to assess the sensitivity of transient calculations to the interphase drag modeling in the core. No judgments were made for the phenomena of phase separation without mixture level formation, mixture level and entrainment in the steam generators, pool formation in the upper plenum, or flow stratification in horizontal pipes because either the phenomenon did not occur in the test or there were insufficient measurements to characterize the behavior.

  9. Simulation experiments of small break LOCA in upper plenum joint pipe for 5 MW heating test reactor

    International Nuclear Information System (INIS)

    Bo Jinhai; Jiang Shengyao; Zhang Youjie; Tong Yunxian; Sun Shusen; Yao Meisheng

    1988-12-01

    A simulation experiment of small break LOCA is introduced, which was caused by the breakdown of a small size or middle size pipe located at upper plenum, or by unexpected opening the safety valve. In the tests, the system pressure, temperature, void fraction and total loss of water were studied. The results showed that the total loss of water was nearly 20% of initial loading water. It means under this condition the 5MW low temperature heating reactor being built in Institute of Nnclear Engergy Technology of Tsinghua University is safe

  10. Application of mesh free lattice Boltzmann method to the analysis of very high temperature reactor lower plenum

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Woon [Dongguk Univ., Gyeongju (Korea, Republic of). Dept. of Energy and Environment

    2011-11-15

    Inside a helium-cooled very high temperature reactor (VHTR) lower plenum, hot gas jets from upper fuel channels with very high velocities and temperatures and is mixed before flowing out. One of the major concerns is local hot spots in the plenum due to inefficient mixing of the helium exiting from differentially heated fuel channels and it involves complex fluid flow physics. For this situation, mesh-free technique, especially Lattice Boltzmann Method (LBM), is thus of particular interest owing to its merit of no mesh generation. As an attempt to find efficiency of the method in such a problem, 3 dimensional flow field inside a scaled test model of the VHTR lower plenum is computed with commercial XFLOW code. Large eddy simulation (LES) and classical Smagorinsky eddy viscosity (EV) turbulence models are employed to investigate the capability of the LBM in capturing large scale vortex shedding. (orig.)

  11. Flow distribution in the inlet plenum of steam generator

    International Nuclear Information System (INIS)

    Khadamakar, H.P.; Patwardhan, A.W.; Padmakumar, G.; Vaidyanathan, G.

    2011-01-01

    Highlights: → Various flow distribution devices have been studied to make the flow distribution uniform in axial as well as tangential direction. → Experiments were performed using Ultrasonic Velocity Profiler (UVP) and Particle Image Velocimetry (PIV). → CFD modeling has been carried out to give more insights. → Various flow distribution devices have been compared. - Abstract: The flow distribution in a 1/5th and 1/8th scale models of inlet plenum of steam generator (SG) has been studied by a combination of experiments and Computational Fluid Dynamics (CFD) simulations. The distribution of liquid sodium in the inlet plenum of the SG strongly affects the thermal as well as mechanical performance of the steam generator. Various flow distribution devices have been used to make the flow distribution uniform in axial as well as tangential direction in the window region. Experiments have been conducted to measure the radial velocity distribution using Ultrasonic Velocity Profiler (UVP) and Particle Image Velocimetry (PIV) under a variety of conditions. CFD modeling has been carried out for various configurations to give more insight into the flow distribution phenomena. The various flow distribution devices have been compared on the basis of a non-uniformity index parameter.

  12. Development of whole core thermal-hydraulic analysis program ACT. 4. Incorporation of three-dimensional upper plenum model

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki

    2003-03-01

    The thermal-hydraulic analysis computer program ACT is under development for the evaluation of detailed flow and temperature fields in a core region of fast breeder reactors under various operation conditions. The purpose of this program development is to contribute not only to clarifying thermal hydraulic characteristics that cannot be revealed by experiments due to measurement difficulty but also to performing rational safety design and assessment. This report describes the incorporation of a three-dimensional upper plenum model to ACT and its verification study as part of the program development. To treat the influence of three-dimensional thermal-hydraulic behavior in a upper plenum on the in-core temperature field, the multi-dimensional general purpose thermal-hydraulic analysis program AQUA, which was developed and validated at JNC, was applied as the base of the upper plenum analysis module of ACT. AQUA enables to model the upper plenum configuration including immersed heat exchangers of the direct reactor auxiliary cooling system (DRACS). In coupling core analysis module that consists of the fuel-assembly and the inter-wrapper gap calculation parts with the upper plenum module, different types of computation mesh systems were jointed using the staggered quarter assembly mesh scheme. A coupling algorithm among core, upper plenum and heat transport system modules, which can keep mass, momentum and energy conservation, was developed and optimized in consideration of parallel computing. ACT was applied to analyzing a sodium experiment (PLANDTL-DHX) performed at JNC, which simulated the natural circulation decay heat removal under DRACS operation conditions for the program verification. From the calculation result, the validity of the improved program was confirmed. (author)

  13. Estimated Uncertainties in the Idaho National Laboratory Matched-Index-of-Refraction Lower Plenum Experiment

    International Nuclear Information System (INIS)

    Donald M. McEligot; Hugh M. McIlroy, Jr.; Ryan C. Johnson

    2007-01-01

    The purpose of the fluid dynamics experiments in the MIR (Matched-Index-of-Refraction) flow system at Idaho National Laboratory (INL) is to develop benchmark databases for the assessment of Computational Fluid Dynamics (CFD) solutions of the momentum equations, scalar mixing, and turbulence models for typical Very High Temperature Reactor (VHTR) plenum geometries in the limiting case of negligible buoyancy and constant fluid properties. The experiments use optical techniques, primarily particle image velocimetry (PIV) in the INL MIR flow system. The benefit of the MIR technique is that it permits optical measurements to determine flow characteristics in passages and around objects to be obtained without locating a disturbing transducer in the flow field and without distortion of the optical paths. The objective of the present report is to develop understanding of the magnitudes of experimental uncertainties in the results to be obtained in such experiments. Unheated MIR experiments are first steps when the geometry is complicated. One does not want to use a computational technique, which will not even handle constant properties properly. This report addresses the general background, requirements for benchmark databases, estimation of experimental uncertainties in mean velocities and turbulence quantities, the MIR experiment, PIV uncertainties, positioning uncertainties, and other contributing measurement uncertainties

  14. Hydraulics in the RPV lower-plenum of EPR

    International Nuclear Information System (INIS)

    Barois, G.; Goreaud, N.; Nicaise, N.

    2001-01-01

    The in-core instrumentation penetrations of the European Pressurised water Reactor (EPR) have been removed from RPV-bottom to RPV-head, leaving empty the lower plenum of the RPV (Reactor Pressure Vessel). In a lower plenum with no internal structure, huge vortices may appear, with negative consequences, such as high disturbance of the core inlet flow distribution, and high increase of the RPV pressure loss. FRAMATOME ANP developed a specific Flow Distribution Device (FDD), annular shaped, located in the RPV lower plenum below the core support plate, which prevents huge vortices from appearing and guarantees a satisfying flow distribution at core inlet in normal operating conditions. The design of the FDD has been optimised with a numerical approach, using the 3-D CFD-code STAR-CD, previously qualified on scale mockup tests. The model developed represents the EPR RPV from the cold leg to core inlet. Thus, the flow distribution at core inlet, the mixing between loop-flows upstream core inlet and the pressure loss in the lower plenum can be evaluated. The optimised FDD provides satisfying performances for all these relevant functional items. (author)

  15. PIV Experiments to Measure Flow Phenomena in a Scaled Model of a VHTR Lower Plenum

    Energy Technology Data Exchange (ETDEWEB)

    Hugh M. McIlroy, Jr.; Donald M. McEligot; Richard R. Schultz; Daniel Christensen; Robert J. Pink; Ryan C. Johnson

    2006-09-01

    A report of experimental data collected at the Matched-Index-of-Refraction (MIR) Laboratory in support of contract DE-AC07-05ID14517 and the INL Standard Problem on measurements of flow phenomena occurring in a lower plenum of a typical prismatic VHTR concept reactor to assess CFD code is presented. Background on the experimental setup and procedures is provided along with several samples of data obtained from the 3-D PIV system and an assessment of experimental uncertainty is provided. Data collected in this study include 3-dimensional velocity-field descriptions of the flow in all four inlet jets and the entire lower plenum with inlet jet Reynolds numbers (ReJet) of approximately 4300 and 12,400. These investigations have generated over 2 terabytes of data that has been processed to describe the various velocity components in formats suitable for external release and archived on removable hard disks. The processed data from both experimental studies are available in multi-column text format.

  16. Study on mixed convective flow penetration into subassembly from reactor hot plenum in FBRs

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, J.; Ohshima, H.; Kamide, H.; Ieda, Y. [Power Reactor and Nuclear Fuel Development Corporation, Ibaraki (Japan)

    1995-09-01

    Fundamental experiments using water were carried out in order to reveal the phenomenon of mixed convective flow penetration into subassemblies from a reactor`s upper plenum of fast breeder reactors. This phenomenon appears under a certain natural circulation conditions during the operation of the direct reactor auxiliary cooling system for decay heat removal and might influence the natural circulation head which determines the core flow rate and therefore affects the core coolability. In the experiment, a simplified model which simulates an upper plenum and a subassembly was used and the ultrasonic velocity profile monitor as well as thermocouples were applied for the simultaneous measurement of velocity and temperature distributions in the subassembly. From the measured data, empirical equations related to the penetration flow onset condition and the penetration depth were obtained using relevant parameters which were derived from dimensional analysis.

  17. Numerical investigation of flow characteristics in a prototypical lower plenum of a prismatic VHTR

    International Nuclear Information System (INIS)

    Ying, Alice; Narula, Manmeet; Abdou, Mohamed; Tsai, Peter; Ando, Yuya

    2007-01-01

    The aim of this study is to obtain insights into the flow behavior, as well as to develop predictive capability with regards to the flow and thermal mixing, that occurs in the lower plenum of a typical prismatic VHTR (Very High Temperature Reactor) concept. In this paper, numerical modeling has been used to capture qualitative phenomena observed during an experiment performed at INL, using a finite volume, thermo-fluid solver system, 'SC/Tetra' from CRADLE. The choice of the correct turbulence model is critical to accurately predict the flow in the VHTR lower plenum. Four different turbulence models have been used in this study and the flow predictions are significantly different. A trail of marker particles and fluid temperature as a passive scalar have been used to qualitatively study the flow characteristics, specifically the turbulent mixing of water jets. The quantitative experimental data, when available, will be used to compare and improve on the available turbulence models. Preliminary numerical modeling has been carried out to address the issue of hot streaking and buoyancy effects of hot helium jets in the lower plenum. (author)

  18. Design of a new SI engine intake manifold with variable length plenum

    International Nuclear Information System (INIS)

    Ceviz, M.A.; Akin, M.

    2010-01-01

    This paper investigates the effects of intake plenum length/volume on the performance characteristics of a spark-ignited engine with electronically controlled fuel injectors. Previous work was carried out mainly on the engine with carburetor producing a mixture desirable for combustion and dispatching the mixture to the intake manifold. The more stringent emission legislations have driven engine development towards concepts based on electronic-controlled fuel injection rather than the use of carburetors. In the engine with multipoint fuel injection system using electronically controlled fuel injectors has an intake manifold in which only the air flows and, the fuel is injected onto the intake valve. Since the intake manifolds transport mainly air, the supercharging effects of the variable length intake plenum will be different from carbureted engine. Engine tests have been carried out with the aim of constituting a base study to design a new variable length intake manifold plenum. Engine performance characteristics such as brake torque, brake power, thermal efficiency and specific fuel consumption were taken into consideration to evaluate the effects of the variation in the length of intake plenum. The results showed that the variation in the plenum length causes an improvement on the engine performance characteristics especially on the fuel consumption at high load and low engine speeds which are put forward the system using for urban roads. According to the test results, plenum length must be extended for low engine speeds and shortened as the engine speed increases. A system taking into account the results of the study was developed to adjust the intake plenum length.

  19. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 3. Numerical investigation for thermal stratification phenomena in the upper plenum

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Yamaguchi, Akira

    2002-06-01

    A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermal stratification characteristics in the upper plenum, and to investigate trade-off relations between gas entrainment and thermal stratification phenomena on in-vessel structures for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) Dummy plug insertion to a slit of the upper core structure is one of the effective measures to stabilize the in-vessel flow patterns and to mitigate in-vessel thermal shocks. (2) Though flow guide device such as a baffle ring attached to reactor vessel wall is an effective measure to eliminate impinging jet to dipped plate, rising characteristics of the thermal stratification interface are affected by the baffle ring devise. (3) Thermal stratification characteristics are not influenced very much by the installation of a partial inner barrel to the dipped plate, which is an effective measure to reduce the horizontal flow velocity components at free surface. (4) Labyrinth structures to the gap between the reactor vessel wall and the outer dipped plate have direct effects upon in-vessel thermal shock characteristics including thermal stratification phenomena due to the closing of flow path between the upper plenum and the free surface plenum. (author)

  20. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    International Nuclear Information System (INIS)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor

  1. Effects of lower plenum flow structure on core inlet flow of ABWR

    International Nuclear Information System (INIS)

    Watanabe, Shun; Abe, Yutaka; Kaneko, Akiko; Watanabe, Fumitoshi; Tezuka, Kenichi

    2010-01-01

    The evaluation of coolant flow structure at a lower plenum of an advanced boiling water reactor (ABWR) in which there are many structures is very important in order to improve generating power. Although the simulation results by CFD (Computational Fluid Dynamics) codes can predict such complicated flow in the lower plenum, it is required to establish the database of flow structure in lower plenum of ABWR experimentally for the benchmark of the CFD codes. In the model of the lower plenum, we measured velocity profiles with LDV and PIV. And differential pressure of constructed model is measured with differential pressure instrument. It was identified that the velocity and differential pressure profiles also showed the tendency to be flat in the core inlet. Moreover, vortexes were observed around side entry orifice by PIV measurement. (author)

  2. TMI-2 reactor vessel plenum final lift

    International Nuclear Information System (INIS)

    Wilson, D.C.

    1986-01-01

    Removal of the plenum assembly from the TMI-2 reactor vessel was necessary to gain access to the core region for defueling. The plenum was lifted from the reactor vessel by the polar crane using three specially designed pendant assemblies. It was then transferred in air to the flooded deep end of the refueling canal and lowered onto a storage stand where it will remain throughout the defueling effort. The lift and transfer were successfully accomplished on May 15, 1985 in just under three hours by a lift team located in a shielded area within the reactor building. The success of the program is attributed to extensive mockup and training activities plus thorough preparations to address potential problems. 54 refs

  3. Fundamental validation of simulation method for thermal stratification in upper plenum of fast reactors. Analysis of sodium experiment

    International Nuclear Information System (INIS)

    Ohno, Shuji; Ohshima, Hiroyuki; Sugahara, Akihiro; Ohki, Hiroshi

    2010-01-01

    Three-dimensional thermal-hydraulic analyses have been carried out for a sodium experiment in a relatively simple axis-symmetric geometry using a commercial CFD code in order to validate simulating methods for thermal stratification behavior in an upper plenum of sodium-cooled fast reactor. Detailed comparison between simulated results and experimental measurement has demonstrated that the code reproduced fairly well the fundamental thermal stratification behaviors such as vertical temperature gradient and upward movement of a stratification interface when utilizing high-order discretization scheme and appropriate mesh size. Furthermore, the investigation has clarified the influence of RANS type turbulence models on phenomena predictability; i.e. the standard k-ε model, the RNG k-ε model and the Reynolds Stress Model. (author)

  4. Intake plenum volume and its influence on the engine performance, cyclic variability and emissions

    International Nuclear Information System (INIS)

    Ceviz, M.A.

    2007-01-01

    Intake manifold connects the intake system to the intake valve of the engine and through which air or air-fuel mixture is drawn into the cylinder. Details of the flow in intake manifolds are extremely complex. Recently, most of engine companies are focused on variable intake manifold technology due to their improvement on engine performance. This paper investigates the effects of intake plenum volume variation on engine performance and emissions to constitute a base study for variable intake plenum. Brake and indicated engine performance characteristics, coefficient of variation in indicated mean effective pressure (COV imep ) as an indicator for cyclic variability, pulsating flow pressure in the intake manifold runner, and CO, CO 2 and HC emissions were taken into consideration to evaluate the effects of different plenum volumes. The results of this study showed that the variation in the plenum volume causes an improvement on the engine performance and the pollutant emissions. The brake torque and related performance characteristics improved pronouncedly about between 1700 and 2600 rpm by increasing plenum volume. Additionally, although the increase in the plenum volume caused the mixture leaner due to the increase in the intake runner pressure and lean mixtures inclined to increase the cyclic variability, a decrease was interestingly observed in the COV imep

  5. Initial Scaling Studies and Conceptual Thermal Fluids Experiments for the Prismatic NGNP Point Design

    Energy Technology Data Exchange (ETDEWEB)

    D. M. McEligot; G. E. McCreery

    2004-09-01

    The objective of this report is to document the initial high temperature gas reactor scaling studies and conceptual experiment design for gas flow and heat transfer. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/ATHENA/RELAP5-3D calculations for the same geometry. Two aspects of the complex flow in an NGNP are being addressed: (1) flow and thermal mixing in the lower plenum ("hot streaking" issue) and (2) turbulence and resulting temperature distributions in reactor cooling channels ("hot channel" issue). Current prismatic NGNP concepts are being examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses are being applied to determine key non-dimensional parameters and their magnitudes over this operating range. For normal operation, the flow in the coolant channels can be considered to be dominant forced convection with slight transverse property variation. The flow in the lower plenum can locally be considered to be a situation of multiple buoyant jets into a confined density-stratified crossflow -- with obstructions. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls - with near stagnant surroundings at one extreme and significant crossflow at the other. Two heat transfer experiments are being considered. One addresses the "hot channel" problem, if necessary. The second experiment will treat heated jets entering a model plenum. Unheated MIR (Matched-Index-of-Refraction) experiments are first steps when the geometry is complicated. One does not want to use a computational technique which will not even handle constant properties properly. The MIR experiment will simulate flow features of the paths of jets

  6. Effects of upper plenum injection on thermo-hydrodynamic behavior under refill and reflood phases

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Sobajima, Makoto; Abe, Yutaka; Adachi, Hiromichi; Ohnuki, Akira; Osakabe, Masahiro

    1984-12-01

    In order to investigate the thermo-hydrodynamic behavior in core under simultaneous ECC water injection into the upper plenum and the intact cold leg during the refill and reflood phases of a PWR-LOCA, Tests S1-SH3 and S1-SH4 were performed by using Slab Core Test Facility (SCTF) with the injection of saturated and 67K subcooled water into the upper plenum, respectively, under the same cold leg injection condition. The following major findings were obtained by examining these test results. (1) Although the core was cooled by the fall back water from the upper plenum into the core during the period of high injection rate into the upper plenum, the core was cooled mainly by the bottom flooding after the BOCREC (Bottom of core recovery). (2) The possible fall back flow rate estimated with a CCFL correlation rapidly decreased after the BOCREC because of the increase of steam generation rate in core. (3) Continuous fall back of subcooled water was not observed even under the condition with large upper plenum injection rate of subcooled water and with steam outflow through the lower plenum into the downcomer. The fall back was intermittently limited by the rapid increase of upward steam flow which was generated in the core due to the evaporation of the fall back water. (4) The rising of liquid level in the lower plenum was suppressed by the pressurization in core due to the evaporation of fall back water before the BOCREC and therefore the beginning of bottom reflood was delayed. Some selected data from Tests S1-SH3 and S1-SH4 are also included in this report. (author)

  7. Experimental Modeling of VHTR Plenum Flows during Normal Operation and Pressurized Conduction Cooldown

    Energy Technology Data Exchange (ETDEWEB)

    Glenn E McCreery; Keith G Condie

    2006-09-01

    The Very High Temperature Reactor (VHTR) is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S. which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. The present document addresses experimental modeling of flow and thermal mixing phenomena of importance during normal or reduced power operation and during a loss of forced reactor cooling (pressurized conduction cooldown) scenario. The objectives of the experiments are, 1), provide benchmark data for assessment and improvement of codes proposed for NGNP designs and safety studies, and, 2), obtain a better understanding of related phenomena, behavior and needs. Physical models of VHTR vessel upper and lower plenums which use various working fluids to scale phenomena of interest are described. The models may be used to both simulate natural convection conditions during pressurized conduction cooldown and turbulent lower plenum flow during normal or reduced power operation.

  8. Numerical study of hot-leg ECC injection into the upper plenum of a pressurized water reactor

    International Nuclear Information System (INIS)

    Daly, B.J.; Torrey, M.D.; Rivard, W.C.

    1981-01-01

    In certain pressurized water reactor (PWR) designs, emergency core coolant (ECC) is injected through the hot legs into the upper plenum. The condensation of steam on this subcooled liquid stream reduces the pressure in the hot legs and upper plenum and thereby affects flow conditions throughout the reactor. In the present study, we examine countercurrent steam-water flow in the hot leg to determine the deceleration of the ECC flow that results from an adverse pressure gradient and from momentum exchange from the steam by interfacial drag and condensation. For the parameters examined in the study, water flow reversal is observed for a pressure drop of 22 to 32 mBar over the 1.5 m hot leg. We have also performed a three-dimensional study of subcooled water injection into air and steam environments of the upper plenum. The ECC water is deflected by an array of cylindrical guide tubes in its passage through the upper plenum. Comparisons of the air-water results with data obtained in a full scale experiment shows reasonable agreement, but indicates that there may be too much resistance to horizontal flow about the columns because of the use of a stair-step representation of the cylindrical guide tube cross section. Calculations of flow past single columns of stair-step, square and circular cross section do indicate excessive water deeentrainment by the noncircular column. This has prompted the use of an arbitrary mesh computational procedure to more accuratey represent the circular cross-section guide tubes. 15 figures

  9. Enhancing the x-ray output of a single-wire explosion with a gas-puff based plasma opening switch

    Science.gov (United States)

    Engelbrecht, Joseph T.; Ouart, Nicholas D.; Qi, Niansheng; de Grouchy, Philip W.; Shelkovenko, Tatiana A.; Pikuz, Sergey A.; Banasek, Jacob T.; Potter, William M.; Rocco, Sophia V.; Hammer, David A.; Kusse, Bruce R.; Giuliani, John L.

    2018-02-01

    We present experiments performed on the 1 MA COBRA generator using a low density, annular, gas-puff z-pinch implosion as an opening switch to rapidly transfer a current pulse into a single metal wire on axis. This gas-puff on axial wire configuration was studied for its promise as an opening switch and as a means of enhancing the x-ray output of the wire. We demonstrate that current can be switched from the gas-puff plasma into the wire, and that the timing of the switch can be controlled by the gas-puff plenum backing pressure. X-ray detector measurements indicate that for low plenum pressure Kr or Xe shots with a copper wire, this configuration can offer a significant enhancement in the peak intensity and temporal distribution of radiation in the 1-10 keV range.

  10. Proposed retrofit of HEPA filter plenums with injection and sampling manifolds for in-place filter testing

    Energy Technology Data Exchange (ETDEWEB)

    Fretthold, J.K. [EG& G Rocky Flats, Inc., Golden, CO (United States)

    1995-02-01

    The importance of testing HEPA filter exhaust plenums with consideration for As Low as Reasonably Achievable (ALARA) will require that new technology be applied to existing plenum designs. HEPA filter testing at Rocky Flats has evolved slowly due to a number of reasons. The first plenums were built in the 1950`s, preceding many standards. The plenums were large, which caused air dispersal problems. The systems were variable air flow. Access to the filters was difficult. The test methods became extremely conservative. Changes in methods were difficult to make. The acceptance of new test methods has been made in recent years with the change in plant mission and the emphasis on worker safety.

  11. Potential for HEPA filter damage from water spray systems in filter plenums

    Energy Technology Data Exchange (ETDEWEB)

    Bergman, W. [Lawrence Livermore National Lab., CA (United States); Fretthold, J.K. [Rocky Flats Safe Sites of Colorado, Golden, CO (United States); Slawski, J.W. [Department of Energy, Germantown, MD (United States)

    1997-08-01

    The water spray systems in high efficiency particulate air (HEPA) filter plenums that are used in nearly all Department of Energy (DOE) facilities for protection against fire was designed under the assumption that the HEPA filters would not be damaged by the water sprays. The most likely scenario for filter damage involves filter plugging by the water spray, followed by the fan blowing out the filter medium. A number of controlled laboratory tests that were previously conducted in the late 1980s are reviewed in this paper to provide a technical basis for the potential HEPA filter damage by the water spray system in HEPA filter plenums. In addition to the laboratory tests, the scenario for BEPA filter damage during fires has also occurred in the field. A fire in a four-stage, BEPA filter plenum at Rocky Flats in 1980 caused the first three stages of BEPA filters to blow out of their housing and the fourth stage to severely bow. Details of this recently declassified fire are presented in this paper. Although these previous findings suggest serious potential problems exist with the current water spray system in filter plenums, additional studies are required to confirm unequivocally that DOE`s critical facilities are at risk. 22 refs., 15 figs.

  12. Empirical method to calculate Clinch River Breeder Reactor (CRBR) inlet plenum transient temperatures

    International Nuclear Information System (INIS)

    Howarth, W.L.

    1976-01-01

    Sodium flow enters the CRBR inlet plenum via three loops or inlets. An empirical equation was developed to calculate transient temperatures in the CRBR inlet plenum from known loop flows and temperatures. The constants in the empirical equation were derived from 1/4 scale Inlet Plenum Model tests using water as the test fluid. The sodium temperature distribution was simulated by an electrolyte. Step electrolyte transients at 100 percent model flow were used to calculate the equation constants. Step electrolyte runs at 50 percent and 10 percent flow confirmed that the constants were independent of flow. Also, a transient was tested which varied simultaneously flow rate and electrolyte. Agreement of the test results with the empirical equation results was good which verifies the empirical equation

  13. CFD predictions of standby liquid control system mixing in lower plenum of a BWR

    International Nuclear Information System (INIS)

    Boyd, Christopher; Skarda, Raymond

    2014-01-01

    Highlights: • Computational fluid dynamics analysis of BWR lower plenum. • Mixing and stratification of the standby liquid control system injection. • Scoping study highlights the expected flow paths and limitations of experiments. - Abstract: During an anticipated transient without scram (ATWS) scenario in certain boiling water reactor (BWR) systems, a standby liquid control system (SLCS) is used to inject a sodium pentaborate solution into the reactor system in order to quickly shut down (scram) the reactor without the use of the control rods. Some BWR designs utilize a SLCS that injects through a set of nozzles on a vertical pipe in the peripheral region of the lower plenum of the reactor vessel. During the scenario, system water levels are reduced and natural circulation flow rates down through the jet pump nozzles and up into the core are a small fraction of the rated system flow. It is during this period that the SLCS flows are considered. This work outlines some initial scoping studies completed by the staff at the Nuclear Regulatory Commission (NRC). An attempt at benchmarking the computational fluid dynamics (CFD) approach using a set of available test data from a small facility is outlined. Due to our lack of information related to specific details of the facility geometry along with the limited data available from the test, the benchmark exercise produced only a qualitative basis for selecting turbulence models and mesh density. A CFD model simulating a full-scale reactor system is developed for the lower plenum of a representative BWR/4 design and SLCS flows and mixing are studied under a range of flow conditions. The full-scale BWR simulation builds upon the lessons learned from the benchmark exercise. One challenge for this work is the large size of the domain and the relatively small size of the geometric details such as flow passages and gaps. The geometry is simplified to make meshing feasible by eliminating some of the small features. The

  14. Experimental Measurement of Flow Phenomena in a VHTR Lower Plenum Model

    Energy Technology Data Exchange (ETDEWEB)

    Hugh M. McIlroy Jr.; Keith G. Condie; Glenn E. McCreery; Donald M. McEligot; Robert J. Pink

    2006-06-01

    The Very-High-Temperature Reactor (VHTR) is one of six reactor technologies chosen for further development by the Generation IV International Forum. In addition this system is the leading candidate for the Next Generation Nuclear Power (NGNP) Project in the U.S which has the goal of demonstrating the production of emissions free electricity and hydrogen by 2015. In preparation for the thermal-hydraulics and safety analyses that will be required to confirm the performance of the NGNP, work has begun on readying the computational tools that will be needed to predict the thermal-hydraulics conditions and safety margins of the reactor design. Meaningful feasibility studies for VHTR designs will require accurate, reliable predictions of material temperatures which depend upon the thermal convection in the coolant channels of the core and other components. Unfortunately, one-dimensional system codes for gas-cooled reactors typically underpredict these temperatures, particularly for reduced power operations and hypothesized accident scenarios. Likewise, most turbulence models in general-purpose CFD codes also underpredict these temperatures. Matched-Index-of-Refraction (MIR) fluid dynamics experiments have been designed and built to develop benchmark databases for the assessment of CFD solutions of the momentum equations, scalar mixing and turbulence models for typical VHTR plenum geometries in the limiting case of negligible buoyancy and constant fluid properties.

  15. Effect of upper plenum water accumuration on reflooding phenomena under forced-feed flooding in SCTF Core-I tests

    International Nuclear Information System (INIS)

    Sudo, Yukio; Sobajima, Makoto; Iwamura, Takamichi; Osakabe, Masahiro; Ohnuki, Akira; Abe, Yutaka; Adachi, Hiromichi

    1983-07-01

    Large Scale Reflood Test Program has been performed under contract with the Atomic Energy Bureau of Science and Technology Agency of Japan since 1976. The Slab Core Test Program is a part of the Large Scale Reflood Test Program along with the Cylindrical Core Test Program. Major purpose of the Slab Core Test Program is to investigate two-dimensional, thermo-hydrodynamic behavior in the core and the effect of fluid communication between the core and the upper plenum on the reflood phenomena in a postulated loss-of-coolant accident of a PWR. A significant upper plenum water accumulation was observed in the Base Case Test Sl-01 which was carried out under forced-feed flooding condition. To investigate the effects of upper plenum water accumulation on reflooding phenomena, accumulated water is extracted out of the upper plenum in Test Sl-03 by full opening of valves for extraction lines located just above the upper core support plate. This report presents this effect of upper plenum water accumulation on reflooding phenomena through the comparison of Tests Sl-01 and Sl-03. In spite of full opening of valves for upper plenum water extraction in Test Sl-03, a little water accumulation was observed which is of the same magnitude as in Test Sl-01 for about 200 s after the beginning of reflood. From 200 s after the beginning of reflood, however, the upper plenum water accumulation is much less in Test Sl-03 than in Test Sl-01, showing the following effects of upper plenum water accumulation. In Test Sl-03, (1) the two-dimensionality of horizontal fluid distribution is much less both above and in the core, (2) water carryover through hot leg and water accumulation in the core are less, (3) quench time is rather delayed in the upper part of the core by less water fall back from the upper plenum, and (4) difference in the core thermal behavior and core heat transfer are not significant in the middle and lower part of the core. (author)

  16. Modeling study of deposition locations in the 291-Z plenum

    International Nuclear Information System (INIS)

    Mahoney, L.A.; Glissmeyer, J.A.

    1994-06-01

    The TEMPEST (Trent and Eyler 1991) and PART5 computer codes were used to predict the probable locations of particle deposition in the suction-side plenum of the 291-Z building in the 200 Area of the Hanford Site, the exhaust fan building for the 234-5Z, 236-Z, and 232-Z buildings in the 200 Area of the Hanford Site. The Tempest code provided velocity fields for the airflow through the plenum. These velocity fields were then used with TEMPEST to provide modeling of near-floor particle concentrations without particle sticking (100% resuspension). The same velocity fields were also used with PART5 to provide modeling of particle deposition with sticking (0% resuspension). Some of the parameters whose importance was tested were particle size, point of injection and exhaust fan configuration

  17. Reduction of sound transmission across plenum windows by incorporating an array of rigid cylinders

    Science.gov (United States)

    Tang, S. K.

    2018-02-01

    The potential improvement of plenum window noise reduction by installing rigid circular cylinder arrays into the window cavity is investigated numerically using the finite-element method in this study. A two-dimensional approach is adopted. The sound transmission characteristics and propagation within the plenum window are also examined in detail. Results show that the installation of the cylinders in general gives rise to broadband improvement of noise reduction across a plenum window regardless of the direction of sound incidence. Such acoustical performance becomes better when more cylinder columns are installed, but it is suggested that the number of cylinder rows should not exceed two. Results also show that the cylinder positions relative to the nodal/anti-nodal planes of the acoustic modes are crucial in the noise reduction enhancement mechanisms. Noise reduction can further be enhanced by staggering the cylinder rows, such that each cylinder row supports the development of a different acoustic mode. For the simple cylinder arrangements considered in this study, the traffic noise reduction enhancement observed in this study can be as high as 4-5 dB, which is already comparable to or higher than the maximum achieved by installing sound absorption into a plenum window.

  18. Experimental study on gas-puff Z-pinch load characteristics on yang accelerator

    International Nuclear Information System (INIS)

    Ren Xiaodong; Huang Xianbin; Yang Libing; Dan Jiakun; Duan Shuchao; Zhang Zhaohui; Zhou Shaotong

    2010-01-01

    A supersonic single-shell gas-puff load has been developed for Z-pinch experiments on 'Yang' accelerator. Using a fast responding pressure probe to measure the supersonic gas flow, impact pressure at different position and plenum pressure were acquired, which were combined with gas dynamics formulas to determine gas pressures and densities. The radial density profile displays that positions of gas shell varies with axial position, and the gas densities on axis increases as the distance from nozzle increases. Integral radial densities indicates that the linear mass density peaks at nozzle exit and decreases as increasing the distance from nozzle. Using single-shell supersonic gas-puff load, Z-pinch implosion experiments were performed on 'Yang' accelerator. Primary analysis of implosion process was presented, and computational trajectories of imploding plasma shell using snowplow model are in agreement with the experimental results. (authors)

  19. Gratiae plenum: Latin, Greek and the Cominform

    Directory of Open Access Journals (Sweden)

    David Movrin

    2010-12-01

    Full Text Available The survival of classics in the People’s Republic of Slovenia after World War II was dominated by the long shadow of the Coryphaeus of the Sciences, Joseph Stalin. Since 1945, the profile of the discipline was determined by the Communist Party, which followed the Soviet example, well-nigh destroying the classical education in the process. Fran Bradač, head of Classics at the University of Ljubljana, was removed for political reasons; the classical gymnasium belonging to the Church was closed down; Greek was struck from the curriculum of the two remaining state classical gymnasia; Latin, previously a central subject at every gymnasium, was severely reduced in 1945, only to disappear entirely in 1946. The classicists who continued to teach were forced to take ‘reorientation courses’ which enabled them to teach Russian and other more suitable subjects. By 1949, only two out of the 42 classicists employed by the Ministry of Education were actually teaching Latin. The Classics department at the university, where only two students were studying in 1949, was on the brink of closure.  Paradoxically, the classical tradition was saved by Stalin’s attack on the same Party. The Cominform conflict in 1948 astonished the Yugoslav communists and pushed them towards a tactical détente with the West, prompting a revision of some of their policies, including education. The process was led by the top echelons of the Party — such as Milovan Djilas, head of the central Agitprop, Boris Kidrič, in charge of Yugoslav economy, and Edvard Kardelj, the Party’s chief ideologue — during the Third Plenum of the Central Committee Politburo in Belgrade in December 1949. Their newly discovered love of Latin and Greek, documented in the minutes of the Politburo Plenum, was overseen only by the discriminating eye of Josip Broz Tito. Classical gymnasia were revived, Latin was reintroduced to some of the other gymnasia, students returned to study classics at the

  20. Summary report of incineration plenum fire: Building 771, July 2, 1980

    International Nuclear Information System (INIS)

    Fretthold, J.K.

    1995-01-01

    At about 1100 on July 2, 1980, a temperature rise above normal was recorded on charts monitoring operation of the incinerator in Room 149, Building 771. The plenum overheat alarm sounded at 1215, emergency actions initiated, and the fire was extinguished and mop-up began at about 1300. Investigation determined that the fire in the plenum was caused by a heat rise in the system, a deteriorated bypass valve on the No. 3 heat exchanger (KOH scrubber), nitration of the urethane seal on the HEPA filter media to the filter frame, and accumulation of metallic fines on the filter media. It was concluded that the management system responded properly, except for the ring- down system to activate the Emergency Operations Center

  1. Development of thermal mixing enhancement method for lower plenum of the High Temperature Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Gradecka, Malwina Joanna, E-mail: malgrad@gmail.com; Woods, Brian G., E-mail: brian.woods@oregonstate.edu

    2016-08-15

    Highlights: • Coolant mixing in lower plenum might be insufficient and pose operational issues. • Two mixing methods were developed to lower the coolant temperature variation. • The methods resulted with reduction of the temperature variation by 60% and 71%. - Abstract: The High Temperature Gas-cooled Reactor (HTGR) is one of the most mature Gen IV reactor concepts under development today. The High Temperature Test Facility (HTTF) at Oregon State University is a test facility that supports the R&D needs for HTGRs. This study focuses on the issue of helium mixing after the core section in the HTTF, the results of which are generally applicable in HTGRs. In the HTTF, hot helium jets at different temperatures are supposed to uniformly mix in the lower plenum (LP) chamber. However, the level of mixing is not sufficient to reduce the peak helium temperature before the hot jet impinges the LP structure, which can cause issues with structural materials and operational issues in the heat exchanger downstream. The maximum allowable temperature variation in the outlet duct connected to the lower plenum is defined as 40 K (±20 K from the average temperature), while the CFD simulations of this study indicate that the reference design suffers temperature variations in the duct as high as 100 K. To solve this issue, the installation of mixing-enhancing structures within the outlet duct were proposed and analyzed using CFD modeling. We show that using either an optimized “Kwiat” structure (developed in this study) or a motionless mixer installed in the outlet duct, the temperature variations can be brought dramatically, with acceptable increases in pressure drop. The optimal solution appears to be to install double motionless mixers with long blades in the outlet duct, which brings the temperature variation into the acceptable range (from 100 K down to 18 K), with a resulting pressure drop increase in the HTTF loop of 0.73 kPa (6% of total pressure drop).

  2. Development of thermal mixing enhancement method for lower plenum of the High Temperature Test Facility

    International Nuclear Information System (INIS)

    Gradecka, Malwina Joanna; Woods, Brian G.

    2016-01-01

    Highlights: • Coolant mixing in lower plenum might be insufficient and pose operational issues. • Two mixing methods were developed to lower the coolant temperature variation. • The methods resulted with reduction of the temperature variation by 60% and 71%. - Abstract: The High Temperature Gas-cooled Reactor (HTGR) is one of the most mature Gen IV reactor concepts under development today. The High Temperature Test Facility (HTTF) at Oregon State University is a test facility that supports the R&D needs for HTGRs. This study focuses on the issue of helium mixing after the core section in the HTTF, the results of which are generally applicable in HTGRs. In the HTTF, hot helium jets at different temperatures are supposed to uniformly mix in the lower plenum (LP) chamber. However, the level of mixing is not sufficient to reduce the peak helium temperature before the hot jet impinges the LP structure, which can cause issues with structural materials and operational issues in the heat exchanger downstream. The maximum allowable temperature variation in the outlet duct connected to the lower plenum is defined as 40 K (±20 K from the average temperature), while the CFD simulations of this study indicate that the reference design suffers temperature variations in the duct as high as 100 K. To solve this issue, the installation of mixing-enhancing structures within the outlet duct were proposed and analyzed using CFD modeling. We show that using either an optimized “Kwiat” structure (developed in this study) or a motionless mixer installed in the outlet duct, the temperature variations can be brought dramatically, with acceptable increases in pressure drop. The optimal solution appears to be to install double motionless mixers with long blades in the outlet duct, which brings the temperature variation into the acceptable range (from 100 K down to 18 K), with a resulting pressure drop increase in the HTTF loop of 0.73 kPa (6% of total pressure drop).

  3. ECC delivery to lower plenum under downcomer injection part 2. RELAP5 assessment

    International Nuclear Information System (INIS)

    Bang, Young Seok; Shin, An Dong; Kim, Hho Jung

    2000-01-01

    In the present study, the capability of the thermal-hydraulic codes, RELAP5/MOD3.2.2 gamma, in predicting the steam-water interaction and the related ECC delivery to lower plenum under downcomer injection condition during refill phase is evaluated using the experimental data of the UPTF Test 21A. The facility is modeled in detail, and the test condition simulated for code calculations. The calculation result is compared with the applicable measurement data and discussed for the pressure response, ECC bypass behavior, lower plenum delivery, global water mass distribution, and local behavior in downcomer

  4. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

    1991-04-01

    The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix ''strike'' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. 5 refs., 5 figs

  5. Two-phase flow in the upper plenum of a boiling water nuclear reactor

    International Nuclear Information System (INIS)

    Tinoco, Hernan

    2003-01-01

    The end part of the Emergency Core Spray System (ECSS) of the Boiling Water Reactors (BWRs) at Forsmark Nuclear Power Plant (NPP) is situated in the Upper Plenum. It consists of a pipe network equipped with water injection nozzles. In case of Lost-of-Coolant Accidents (LOCAs), the ECSS should maintain the core covered by water and, at the same time, rapidly cool and decompress the reactor by means of cold water injection. In similar reactors, some welds belonging to the ECSS support have, after a period of time, shown crack indications. Inspection, repair or replacement of these welds is time consuming and expensive. For this reason, it has now been decided to permanently remove the end part of the ECSS and to replace it by water injection in the Downcomer. However, this removal should not be accompanied by undesirable effects like an increase in the moisture of the steam used for operating the turbines. To investigate the effect of this removal on the steam moisture, a CFD analysis of the two-phase flow in the Upper Plenum of Unit 3, with and without ECSS, has been carried out by means of a two-phase Euler model in FLUENT 6.0. The inlet conditions are given by an analysis of the core kinetics and thermal hydraulics by mean of the POLCA-code. The outlet conditions, i. e. the steam separator pressure drops, are given by empirical correlations from the experiments carried out at the SNORRE facility. The predicted the mass flow-rates to each separator, together with empirical correlations for the moisture content of the steam leaving the separators and the steam dryer, indicate a slight decrease in the steam moisture when the ECSS is removed. Also, a minor decrease in pressure losses over the Upper Plenum is achieved with this removal. On the other hand, rounding the sharp edges of the inlet openings to the steam separators at the shroud cover may give a large reduction in pressure losses

  6. Results from AFWL 230 kJ coaxial plasma gun experiments

    International Nuclear Information System (INIS)

    Hall, D.J.; Baker, W.L.; Beason, J.D.; Clouse, C.J.; Degnan, J.H.; Dietz, D.; Hackett, K.E.; Higgins, P.L.; Holmes, J.L.; Price, D.W.

    1988-01-01

    A coaxial plasma gun has been operated on the AFWL 0.5 MJ capacitor bank. A Marshall valve actuated by an explosive detonator is used to puff hydrogen gas from a small high pressure plenum into the breech of the gun. After a set delay from the explosion the capacitor bank is discharged across the electrodes of the coaxial gun. The operating mode of the gun can be changed by varying the plenum pressure and the firing delay. Over 150 shots have been fired, varying delay, plenum pressure, and initial stored energy. Initial plenum pressures were varied from 250 to 750 psi, and firing delays ranged from 0.8 msec to 2.2 msec. Experiments were conducted at 90, 176, and 230 kJ of initial stored energy (50, 70, adn 80 kV charge). Rogowski coils were used to measure current and magnetic field within the plasma at 25 axial locations along the gun. The coils were installed in grooves on the inner surface of the outer conductor. Signals from the coils were passively integrated. Integrator time constants ranged from 95 to 114 μsec. Time histories of magnetic field profiles are presented. These are used to describe the operating mode of the gun

  7. HOT CELL SYSTEM FOR DETERMINING FISSION GAS RETENTION IN METALLIC FUELS

    Energy Technology Data Exchange (ETDEWEB)

    Sell, D. A.; Baily, C. E.; Malewitz, T. J.; Medvedev, P. G.; Porter, D. L.; Hilton, B. A.

    2016-09-01

    A system has been developed to perform measurements on irradiated, sodium bonded-metallic fuel elements to determine the amount of fission gas retained in the fuel material after release of the gas to the element plenum. During irradiation of metallic fuel elements, most of the fission gas developed is released from the fuel and captured in the gas plenums of the fuel elements. A significant amount of fission gas, however, remains captured in closed porosities which develop in the fuel during irradiation. Additionally, some gas is trapped in open porosity but sealed off from the plenum by frozen bond sodium after the element has cooled in the hot cell. The Retained fission Gas (RFG) system has been designed, tested and implemented to capture and measure the quantity of retained fission gas in characterized cut pieces of sodium bonded metallic fuel. Fuel pieces are loaded into the apparatus along with a prescribed amount of iron powder, which is used to create a relatively low melting, eutectic composition as the iron diffuses into the fuel. The apparatus is sealed, evacuated, and then heated to temperatures in excess of the eutectic melting point. Retained fission gas release is monitored by pressure transducers during the heating phase, thus monitoring for release of fission gas as first the bond sodium melts and then the fuel. A separate hot cell system is used to sample the gas in the apparatus and also characterize the volume of the apparatus thus permitting the calculation of the total fission gas release from the fuel element samples along with analysis of the gas composition.

  8. Nondestructive fission gas release measurement and analysis

    International Nuclear Information System (INIS)

    O'Leary, P.M.; Packard, D.R.

    1993-01-01

    Siemens Power Corporation (SPC) has performed reactor poolside gamma scanning measurements of fuel rods for fission gas release (FGR) detection for more than 10 yr. The measurement system has been previously described. Over the years, the data acquisition system, the method of spectrum analysis, and the means of reducing spectrum interference have been significantly improved. A personal computer (PC)-based multichannel analyzer (MCA) package is used to collect, display, and store high-resolution gamma-ray spectra measured in the fuel rod plenum. A PC spread sheet is used to fit the measured spectra and compute sample count rates after Compton background subtraction. A Zircaloy plenum spacer is often used to reduce positron annihilation interference that can arise from the INCONEL reg-sign plenum spring used in SPC-manufactured fuel rods

  9. Interferometric investigation of turbulently fluctuating temperature in an LMFBR outlet plenum geometry

    International Nuclear Information System (INIS)

    Bennett, R.G.; Golay, M.W.

    1975-01-01

    A novel optical technique is described for the measurement of turbulently fluctuating temperature in a transparent fluid flow. The technique employs a Mach-Zehnder interferometer of extremely short field and a simple photoconductive diode detector. The system produces a nearly linear D.C. electrical analog of the turbulent temperature fluctuations in a small, 1 mm 3 volume. The frequency response extends well above 2500 Hz, and can be improved by the choice of a more sophisticated photodetector. The turbulent sodium mixing in the ANL 1 1 / 15 -scale FFTF outlet plenum is investigated with a scale model outlet mixing plenum, using flows of air. The scale design represents a cross section of the ANL outlet plenum, so that the average recirculating flow inside the test cell is two dimensional. The range of the instrument is 120 0 F above the ambient air temperature. The accuracy is generally +-5 0 F, with most of the error due to noise originating from building vibrations and room noise. The power spectral density of the fluctuating temperature has been observed experimentally at six different stations in the flow. A strong 300 Hz component is generated in the inlet region, which decays as the flow progresses along streamlines. The effect of the inlet Reynolds number and the temperature difference between the inlet flows on the power spectral density has also been investigated. Traces of the actual fluctuating temperature are included for the six stations

  10. ISOTHERMAL AIR INGRESS VALIDATION EXPERIMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Chang H Oh; Eung S Kim

    2011-09-01

    Idaho National Laboratory carried out air ingress experiments as part of validating computational fluid dynamics (CFD) calculations. An isothermal test loop was designed and set to understand the stratified-flow phenomenon, which is important as the initial air flow into the lower plenum of the very high temperature gas cooled reactor (VHTR) when a large break loss-of-coolant accident occurs. The unique flow characteristics were focused on the VHTR air-ingress accident, in particular, the flow visualization of the stratified flow in the inlet pipe to the vessel lower plenum of the General Atomic’s Gas Turbine-Modular Helium Reactor (GT-MHR). Brine and sucrose were used as heavy fluids, and water was used to represent a light fluid, which mimics a counter current flow due to the density difference between the stimulant fluids. The density ratios were changed between 0.87 and 0.98. This experiment clearly showed that a stratified flow between simulant fluids was established even for very small density differences. The CFD calculations were compared with experimental data. A grid sensitivity study on CFD models was also performed using the Richardson extrapolation and the grid convergence index method for the numerical accuracy of CFD calculations . As a result, the calculated current speed showed very good agreement with the experimental data, indicating that the current CFD methods are suitable for predicting density gradient stratified flow phenomena in the air-ingress accident.

  11. Modelling of fission gas release in rods from the International DEMO-RAMP-II Project at Studsvik

    International Nuclear Information System (INIS)

    Malen, K.

    1983-01-01

    The DEMO-RAMP-II rods had a burn-up of 25-30 MWd/kg U. They were ramped to powers in the range 40-50 kW/m with hold times between 10 s and 4.5 minutes. In spite of the short hold times the fission gas release at the higher powers was more than 1%. With these short hold times it is natural to assume that mixing of released gas with plenum gas is limited. Modelling has been performed using GAPCONSV (a modified GAPCON-THERMAL-2) both with and without mixing of released gas with plenum gas. In particular for the high power-short duration ramps only the ''no mixing'' modelling yields release fractions comparable to the experimental values. (author)

  12. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

    1991-01-01

    The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. (In the past DIII-D operated with an open divertor.) The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix 'strike' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. (author) 5 refs., 5 figs

  13. Melt jet fragmentation and oxidation in the lower plenum

    International Nuclear Information System (INIS)

    Berthoud, G.

    2001-01-01

    During the late phases of a PWR Severe Accident, the core materials discharge into the lower plenum in which water is still present. In that case, we are then concerned by the possible occurrence of a Steam Explosion which may endanger the vessel structure and by the following cooling of the melt debris. So, we have two possible ways of vessel rupture: a mechanical one following an energetic Steam Explosion and a thermal one due to insufficient debris cooling. Both types of problems are linked with the degree of fragmentation of the core material during its penetration into the water of the lower plenum. One of the most likely mode of discharge consists in corium streams or jets. The fragmentation will build a corium-water mixture (the pre-mixing sequence) which, under certain circumstances, may undergo a fine fragmentation sequence leading to an energetic Steam Explosion (the explosion sequence). Whatever the occurrence of a Steam Explosion, the resulting debris will accumulate at the bottom of the Reactor Vessel and the cooling of such a ''debris bed'' is known to be highly dependant of the granulometry and build up of the debris bed which are linked with the previous sequence of corium fragmentation and dispersion. In CEA, the MC3D Code has been developed to deal with all these phenomena. (author)

  14. Measurement of heat and momentum eddy diffusivities in recirculating LMFBR outlet plenum flows

    International Nuclear Information System (INIS)

    Manno, V.P.; Golay, M.W.

    1978-06-01

    An optical technique has been developed for the measurement of the eddy diffusivity of heat in a transparent flowing medium. The method uses a combination of two established measurement tools: a Mach-Zehnder interferometer for the monitoring of turbulently fluctuating temperature and a Laser Doppler Anemometer (LDA) for the measurement of turbulent velocity fluctuations. The technique is applied to the investigation of flow fields characteristic of the LMFBR outlet plenum. The study is accomplished using air as the working fluid in a small scale Plexiglas test section. Lows are introduced into both the 1 / 15 scale FFTF outlet plenum and the 3 / 80 scale CRBR geometry plenum at inlet Reynolds numbers of 22,000. Measurements of the eddy diffusivity of heat and the eddy diffusivity of momentum are performed at a total of 11 measurement stations. Significant differences of the turbulence parameters are found between the two geometries, and the higher chimney structure of the CRBR case is found to be the major cause of the distinction. Spectral intensity studies of the fluctuating electronic analog signals of velocity and temperature are also performed. Error analysis of the overall technique indicates an experimental error of 10% in the determination of the eddy diffusivity of heat and 6% in the evaluation of turbulent momentum viscosity. In general it is seen that the turbulence in the cases observed is not isotropic, and use of isotropic turbulent heat and momentum diffusivities in transport modelling would not be a valid procedure

  15. Experimental study on breakup and fragmentation behavior of molten material jet in complicated structure of BWR lower plenum

    International Nuclear Information System (INIS)

    Saito, Ryusuke; Abe, Yutaka; Yoshida, Hiroyuki

    2014-01-01

    To estimate the state of reactor pressure vessel of Fukushima Daiichi nuclear power plant, it is important to clarify the breakup and fragmentation of molten material jet in the lower plenum of boiling water reactor (BWR) by a numerical simulation. To clarify the effects of complicated structures on the jet behavior experimentally and validate the simulation code, we conduct the visualized experiments simulating the severe accident in the BWR lower plenum. In this study, jet breakup, fragmentation and surrounding velocity profiles of the jet were observed by the backlight method and the particle image velocimetry (PIV) method. From experimental results using the backlight method, it was clarified that jet tip velocity depends on the conditions whether complicated structures exist or not and also clarified that the structures prevent the core of the jet from expanding. From measurements by the PIV method, the surrounding velocity profiles of the jet in the complicated structures were relatively larger than the condition without structure. Finally, fragment diameters measured in the present study well agree with the theory suggested by Kataoka and Ishii by changing the coefficient term. Thus, it was suggested that the fragmentation mechanism was mainly controlled by shearing stress. (author)

  16. Large Eddy Simulation of Fluid flow and Heat Transfer in the Upper Plenum of Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Seokki; Lee, Taeho; Kim, Dongeun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ko, Sungho [Chungnam National Univ., Daejeon (Korea, Republic of)

    2014-05-15

    The important parameters in the thermal striping are the frequency and the amplitude of the temperature fluctuation. Since the sodium used as coolant in the PGSFR has a high thermal conductivity, the temperature fluctuation can be easily transferred to the solid walls of the components in the upper plenum. To remedy these problems, numerical studies are performed in the present study to analyze the thermal striping for possible improvement of the design and safety of the reactor. For the numerical works, Chacko et al. performed LES for the experiment by Nam and Kim, and found that the LES can produce the oscillation of temperature fluctuation properly, while the realizable k - ε model predicts the amplitude and frequency of the temperature fluctuation very poorly indicating that the LES method is an appropriate calculation method for the thermal striping. In this paper, the simulation of thermal striping in the upper plenum of PGSFR is performed using the LES method. The WALE eddy viscosity model by Nicoud and Ducros built in CFX-13 commercial code is employed for the LES eddy viscosity model. The numerical investigation of the thermal striping is performed with the LES method using the CFX-13 commercial code, where the solution domain is the upper plenum of the PGSFR. As the first step, dozens of monitoring points are set to locations that are anticipated to cause thermal striping. Then, the temperature fluctuations were calculated along with the time-averaged variables such as the velocity and temperature. From these results we have obtained the following conclusions. At the side wall of IHX, a slight fluctuation is observed, but it seems that there is no risk of thermal striping. The flows from the reactor core are not mixed when reaching the UIS. So both the first and second plates need to be considered. Among the first grid plate regions, the shape region is the weakest region for thermal striping. The second weakest region for thermal striping is the shape

  17. Decontamination and demolition of a former plutonium processing facility's process exhaust system, firescreen, and filter plenum buildings

    International Nuclear Information System (INIS)

    LaFrate, P.J. Jr.; Stout, D.S.; Elliott, J.W.

    1996-01-01

    The Los Alamos National Laboratory (LANL) Decommissioning Project has decontaminated, demolished, and decommissioned a process exhaust system, two filter plenum buildings, and a firescreen plenum structure at Technical Area 21 (TA-2 1). The project began in August 1995 and was completed in January 1996. These high-efficiency particulate air (HEPA) filter plenums and associated ventilation ductwork provided process exhaust to fume hoods and glove boxes in TA-21 Buildings 2 through 5 when these buildings were active plutonium and uranium processing and research facilities. This paper summarizes the history of TA-21 plutonium and uranium processing and research activities and provides a detailed discussion of integrated work process controls, characterize-as-you-go methodology, unique engineering controls, decontamination techniques, demolition methodology, waste minimization, and volume reduction. Also presented in detail are the challenges facing the LANL Decommissioning Project to safely and economically decontaminate and demolish surplus facilities and the unique solutions to tough problems. This paper also shows the effectiveness of the integrated work package concept to control work through all phases

  18. Decontamination and demolition of a former plutonium processing facility's process exhaust system, firescreen, and filter plenum buildings

    International Nuclear Information System (INIS)

    LaFrate, P.J. Jr.; Stout, D.S.; Elliott, J.W.

    1996-01-01

    The Los Alamos National Laboratory (LANL) Decommissioning Project has decontaminated, demolished, and decommissioned a process exhaust system, two filter plenum buildings, and a firescreen plenum structure at Technical Area 21 (TA-21). The project began in August 1995 and was completed in January 1996. These high-efficiency particulate air (HEPA) filter plenums and associated ventilation ductwork provided process exhaust to fume hoods and glove boxes in TA-21 Buildings 2 through 5 when these buildings were active plutonium and uranium processing and research facilities. This paper summarizes the history of TA-21 plutonium and uranium processing and research activities and provides a detailed discussion of integrated work process controls, characterize-as-you-go methodology, unique engineering controls, decontamination techniques, demolition methodology, waste minimization, and volume reduction. Also presented in detail are the challenges facing the LANL Decommissioning Project to safely and economically decontaminate and demolish surplus facilities and the unique solutions to tough problems. This paper also shows the effectiveness of the integrated work package concept to control work through all phases

  19. Operational experience with Dragon reactor experiment of relevance to commercial reactors

    International Nuclear Information System (INIS)

    Capp, P.D.; Simon, R.A.

    1976-01-01

    An important part of the experience gained during the first ten years of successful power operation of the Dragon Reactor is relevant to the design and operation of future High Temperature Reactors (HTRs). The aspects presented in this paper have been chosen as being particularly applicable to larger HTR systems. Core performance under a variety of conditions is surveyed with particular emphasis on a technique developed for the identification and location of unpurged releasing fuel and the presence of activation and fission products in the core area. The lessons learned during the reflector block replacement are presented. Operating experience with the primary circuit identifies the lack of mixing of gas streams within the hot plenum and the problems of gas streaming in ducts. Helium leakage from the circuit is often greater than the optimum 0.1%/d. Virtually all the leakage problems are associated with the small bore instrument pipework essential for the many experiments associated with the Dragon Reactor Experiment (DRE). Primary circuit maintenance work confirms the generally clean state of the DRE circuit but identifies 137 Cs and 110 Agsup(m) as possible hazards if fuel emitting these isotopes is irradiated. (author)

  20. Study of gas-puff Z-pinches on COBRA

    Energy Technology Data Exchange (ETDEWEB)

    Qi, N.; Rosenberg, E. W.; Gourdain, P. A.; Grouchy, P. W. L. de; Kusse, B. R.; Hammer, D. A.; Bell, K. S.; Shelkovenko, T. A.; Potter, W. M.; Atoyan, L.; Cahill, A. D.; Evans, M.; Greenly, J. B.; Hoyt, C. L.; Pikuz, S. A.; Schrafel, P. C. [Laboratory of Plasma Studies, Cornell University, Ithaca, New York 14853 (United States); Kroupp, E.; Fisher, A.; Maron, Y. [Weizmann Institute of Science, Rehovot 76100 (Israel)

    2014-11-15

    Gas-puff Z-pinch experiments were conducted on the 1 MA, 200 ns pulse duration Cornell Beam Research Accelerator (COBRA) pulsed power generator in order to achieve an understanding of the dynamics and instability development in the imploding and stagnating plasma. The triple-nozzle gas-puff valve, pre-ionizer, and load hardware are described. Specific diagnostics for the gas-puff experiments, including a Planar Laser Induced Fluorescence system for measuring the radial neutral density profiles along with a Laser Shearing Interferometer and Laser Wavefront Analyzer for electron density measurements, are also described. The results of a series of experiments using two annular argon (Ar) and/or neon (Ne) gas shells (puff-on-puff) with or without an on- (or near-) axis wire are presented. For all of these experiments, plenum pressures were adjusted to hold the radial mass density profile as similar as possible. Initial implosion stability studies were performed using various combinations of the heavier (Ar) and lighter (Ne) gasses. Implosions with Ne in the outer shell and Ar in the inner were more stable than the opposite arrangement. Current waveforms can be adjusted on COBRA and it was found that the particular shape of the 200 ns current pulse affected on the duration and diameter of the stagnated pinched column and the x-ray yield.

  1. Off-gas characteristics of defense waste vitrification using liquid-fed Joule-heated ceramic melters

    International Nuclear Information System (INIS)

    Goles, R.W.; Sevigny, G.J.

    1983-09-01

    Off-gas and effluent characterization studies have been established as part of a PNL Liquid-Fed Ceramic Melter development program supporting the Savannah River Laboratory Defense Waste Processing Facility (SRL-DWPF). The objectives of these studies were to characterize the gaseous and airborne emission properties of liquid-fed joule-heated melters as a function of melter operational parameters and feed composition. All areas of off-gas interest and concern including effluent characterization, emission control, flow rate behavior and corrosion effects have been studied using alkaline and formic-acid based feed compositions. In addition, the behavioral patterns of gaseous emissions, the characteristics of melter-generated aerosols and the nature and magnitude of melter effluent losses have been established under a variety of feeding conditions with and without the use of auxiliary plenum heaters. The results of these studies have shown that particulate emissions are responsible for most radiologically important melter effluent losses. Melter-generated gases have been found to be potentially flammable as well as corrosive. Hydrogen and carbon monoxide present the greatest flammability hazard of the combustibles produced. Melter emissions of acidic volatile compounds of sulfur and the halogens have been responsible for extensive corrosion observed in melter plenums and in associated off-gas lines and processing equipment. The use of auxiliary plenum heating has had little effect upon melter off-gas characteristics other than reducing the concentrations of combustibles

  2. Analysis research on mixing characteristics of lower plenum of Qinshan phase Ⅱ NPP by CFD method

    International Nuclear Information System (INIS)

    Mao Huihui; He Peifeng; Lu Chuan; Zhang Hongliang

    2015-01-01

    The flowing and mixing characteristics of the lower plenum of Qinshan Phase n NPP were analyzed by CFD method. The calculation results were compared with the results of the reactor hydraulic simulation test. On core inlet mass flow distributions, both upwind and high resolution advection schemes show good agreements with test results. While on lower plenum mixing characteristics, the calculation results from either upwind or high resolution advection schemes show relatively large differences to the test data. Relatively, upwind advection schemes predict better anticipations on maximum and minimum mixing factors. Furthermore, whether or not considering helix flow by main pump is the most possible key factor that leads to difference between CFD calculation and test results. (authors)

  3. Validation Studies for Numerical Simulations of Flow Phenomena Expected in the Lower Plenum of a Prismatic VHTR Reference Design

    International Nuclear Information System (INIS)

    Richard W. Johnson

    2005-01-01

    The final design of the very high temperature reactor (VHTR) of the fourth generation of nuclear power plants (Gen IV) has not yet been established. The VHTR may be either a prismatic (block) or pebble bed type. It may be either gas-cooled or cooled with an as yet unspecified molten salt. However, a conceptual design of a gas-cooled VHTR, based on the General Atomics GT-MHR, does exist and is called the prismatic VHTR reference design, MacDonald et al [2003], General Atomics [1996]. The present validation studies are based on the prismatic VHTR reference design. In the prismatic VHTR reference design, the flow in the lower plenum will be introduced by dozens of turbulent jets issuing into a large crossflow that must negotiate dozens of cylindrical support columns as it flows toward the exit duct of the reactor vessel. The jets will not all be at the same temperature due to the radial variation of power density expected in the core. However, it is important that the coolant be well mixed when it enters the power conversion unit to ensure proper operation and long life of the power conversion machinery. Hence, it is deemed important to be able to accurately model the flow and mixing of the variable temperature coolant in the lower plenum and exit duct. Accurate flow modeling involves determining modeling strategies including the fineness of the grid needed, iterative convergence tolerance, numerical discretization method used, whether the flow is steady or unsteady, and the turbulence model and wall treatment employed. It also involves validation of the computer code and turbulence model against a series of separate and combined flow phenomena and selection of the data used for the validation. The present report describes progress made to date for the task entitled ''CFD software validation of jets in crossflow'' which was designed to investigate the issues pertaining to the validation process

  4. TFTR neutral beam D-T gas injection system operational experiences of the first two years

    International Nuclear Information System (INIS)

    Oldaker, M.E.; Lawson, J.E.; Stevenson, T.N.; Kamperschroer, J.H.

    1995-01-01

    The TFTR Neutral Beam Tritium Gas Injection System (TGIS) has successfully performed tritium operations since December 1993. TGIS operation has been reliable, with no leaks to the secondary containment to date. Notable operational problems include throughput leaks on fill, exit and piezoelectric valves. Repair of a TGIS requires replacement of the assembly, involving TFTR downtime and extensive purging, since the TGIS assembly is highly contaminated with residual tritium, and is located within secondary containment. Modifications to improve reliability and operating range include adjustable reverse bias voltage to the piezoelectric valves, timing and error calculation changes to tune the PLC and hardwired timing control, and exercising piezoelectric valves without actually pulsing gas prior to use after extended inactivity. A pressure sensor failure required the development of an open loop piezoelectric valve drive control scheme, using a simple voltage ramp to partially compensate for declining plenum pressure. Replacement of TGIS's have been performed, maintaining twelve system tritium capability as part of scheduled project maintenance activity

  5. Molten material relocation into the lower plenum: a status report

    International Nuclear Information System (INIS)

    1998-09-01

    This report, prepared by the task group 'Degraded Core Cooling' (DCC) for the CSNI, summarizes the experimental and theoretical knowledge of molten material relocation from a degraded core to the lower plenum of the reactor vessel under the main severe accident scenarios envisaged for both PWRs and BWRs, and boundary conditions. Consequences of movement of material to the lower head are considered with respect to the potential for reactor pressure vessel failure. The following models are reviewed: SCDAP/RELAP5, ICARE/CATHARE, ATHLET-CD/KESS, MELCOR, MAAP4, ESCADRE, etc.

  6. CATHARE2 analysis on the loss of residual heat removal system during mid-loop operation : pressurizer and SGI outlet plenum manways open

    International Nuclear Information System (INIS)

    Chung, Young Jong; Chang, Won Pyo.

    1997-06-01

    The present study is to analyze the BETHSY test 6.9c using CATHARE2 v1.3u. BETHSY test 6.9c simulates plant conditions following loss of residual heat removal system under mid-loop operation. The configuration is that the pressurizer and steam generator outlet plenum manways are opened as vent paths in order to protect the system from overpressurization by removing the steam generated in the core. Most of the important physical phenomena are observed in the experiment have been predicted reasonably by the CATHARE2 code. Since the differential pressure between the pressurizer and the surge line is overestimated, the peak pressure in the upper plenum is predicted higher than the experimental value by 11 kPa and occurrence is delayed by 210s. Also earlier core uncovery is predicted, mainly due to overprediction of the manway flows. The analysis results are demonstrated that opening of the pressurizer and the steam generator outlet plenum manways is effective to prevent the core uncovery by only gravity feed injection. Although some disagreements found in detailed phenomena, the prediction of the overall system behavior by the code does not deviate from the experimental results unacceptably. The core bypass flowrate is found to be very sensitive to mass distribution in the core and the system behaviors are strongly affected by phase separation modeling under low pressure and particularly stratified flow condition. the main purpose of the present study is to understand physical phenomena under the accident and to assess the capability of CATHARE2 prediction for enhancement of reliability in actual plant analyses. (author). 11 refs., 3 tabs., 41 figs

  7. Fracture mechanics evaluation of LOFT lower plenum injection nozzle

    International Nuclear Information System (INIS)

    Nagata, P.K.; Reuter, W.G.

    1977-01-01

    An analysis to establish whether or not a leak-before-break concept would apply to the LOFT lower plenum injection nozzle is described. The analysis encompassed the structure from the inlet side of valve V-2170 to the lower plenum nozzle-to-reactor vessel weld on the left side of the emergency core cooling system (ECCS). The defect that was assumed to exist was of such a size that the probability of its being missed by the applicable inspection technique was near zero. The Inconel 600 nozzle forging with an initial assumed defect size of 0.64 cm (0.25 in.) deep would behave as follows: (1) the axially oriented defect would result in leak before rupture (the number of cycles to rupture was 11,000), (2) the circumferentially oriented defect would result in a rupture before leak. The number of cycles to failure would be in excess of 14,000. Based on the conservative assumption that the thermal stresses were membrane stresses as opposed to a bending stress, the following were found. For the Inconel 82 weld metal (thickness of 1.3 cm [0.53 in.]) and AISI 316 SST valve body, with an initial assumed defect of 0.25 cm (0.1 in.), the crack would grow through the thickness in a minimum of 3950 cycles and to a critical rupture crack length of 5.1 cm (2.0 in.) in an additional 80 cycles. The Inconel 82 weld metal at the shell body (thickness of 9.7 cm or 3.8 in.) with an assumed defect 1.3 cm (0.5 in.) deep would fail in 334 cycles. Calculations made assuming a linear stress gradient instead of the above-mentioned flat distribution through the wall indicated that the number of stress cycles increased to 2200

  8. Study of thermal-hydraulic characteristics in an LMFBR intermediate plenum

    International Nuclear Information System (INIS)

    Uotani, M.; Naohara, N.; Kinoshita, I.

    1985-01-01

    Experimental studies using water and liquid metal were conducted in order to investigate the thermal-hydraulic characteristics of an LMFBR intermediate plenum. The present study is an attempt to evaluate the effect of natural convection on the temperature field and to validate the prediction method of temperature profile in a thermally stratified cavity. The experimental results indicated that the effect of the natural convection on flow velocity and heat transfer in the cavity is reduced with increasing the modified stratification parameter. The calculation by FEM code and a simple 1-D model are effective to predict the temperature profile in the cavity

  9. Advanced technology for aero gas turbine components

    Energy Technology Data Exchange (ETDEWEB)

    1987-09-01

    The Symposium is aimed at highlighting the development of advanced components for new aero gas turbine propulsion systems in order to provide engineers and scientists with a forum to discuss recent progress in these technologies and to identify requirements for future research. Axial flow compressors, the operation of gas turbine engines in dust laden atmospheres, turbine engine design, blade cooling, unsteady gas flow through the stator and rotor of a turbomachine, gear systems for advanced turboprops, transonic blade design and the development of a plenum chamber burner system for an advanced VTOL engine are among the topics discussed.

  10. Investigation of the coolability of a continuous mass of relocated debris to a water-filled lower plenum. Technical report

    International Nuclear Information System (INIS)

    Rempe, J.L.; Wolf, J.R.; Chavez, S.A.; Condie, K.G.; Hagrman, D.L.; Carmack, W.J.

    1994-09-01

    This report documents work performed to support the development of an analytical and experimental program to investigate the coolability of a continuous mass of debris that relocates to a water-filled lower plenum. The objective of this program is to provide an adequate data base for developing and validating a model to predict the coolability of a continuous mass of debris relocating to a water-filled lower plenum. The model must address higher pressure scenarios, such as the TMI-2 accident, and lower pressure scenarios, which recent calculations indicate are more likely for most operating LWR plants. The model must also address a range of possible debris compositions

  11. Scaling studies and conceptual experiment designs for NGNP CFD assessment

    Energy Technology Data Exchange (ETDEWEB)

    D. M. McEligot; G. E. McCreery

    2004-11-01

    The objective of this report is to document scaling studies and conceptual designs for flow and heat transfer experiments intended to assess CFD codes and their turbulence models proposed for application to prismatic NGNP concepts. The general approach of the project is to develop new benchmark experiments for assessment in parallel with CFD and coupled CFD/systems code calculations for the same geometry. Two aspects of the complex flow in an NGNP are being addressed: (1) flow and thermal mixing in the lower plenum ("hot streaking" issue) and (2) turbulence and resulting temperature distributions in reactor cooling channels ("hot channel" issue). Current prismatic NGNP concepts are being examined to identify their proposed flow conditions and geometries over the range from normal operation to decay heat removal in a pressurized cooldown. Approximate analyses have been applied to determine key non-dimensional parameters and their magnitudes over this operating range. For normal operation, the flow in the coolant channels can be considered to be dominant turbulent forced convection with slight transverse property variation. In a pressurized cooldown (LOFA) simulation, the flow quickly becomes laminar with some possible buoyancy influences. The flow in the lower plenum can locally be considered to be a situation of multiple hot jets into a confined crossflow -- with obstructions. Flow is expected to be turbulent with momentumdominated turbulent jets entering; buoyancy influences are estimated to be negligible in normal full power operation. Experiments are needed for the combined features of the lower plenum flows. Missing from the typical jet experiments available are interactions with nearby circular posts and with vertical posts in the vicinity of vertical walls - with near stagnant surroundings at one extreme and significant crossflow at the other. Two types of heat transfer experiments are being considered. One addresses the "hot channel" problem, if necessary

  12. Deformation and collapse of zircaloy fuel rod cladding into plenum axial gaps

    International Nuclear Information System (INIS)

    Pfennigwerth, P.L.; Gorscak, D.A.; Selsley, I.A.

    1983-01-01

    To minimize support structure, blanket and reflector fuel rods of the thoria urania-fueled Light Water Breeder Reactor (LWBR) were designed with non-freestanding Zircaloy-4 cladding. An analytical model was developed to predict deformation of unirradiated cladding into axial gaps of fuel rod plenum regions where it is unsupported. This model uses the ACCEPT finite element computer program to calculate elastic-plastic deformation of cladding due to external pressure. The finite element is 20-node, triquadratic, isoparametric, and 3-dimensional. Its curved surface permits accurate modeling of the tube geometry, including geometric nonuniformities such as circumferential wall thickness variation and initial tube out-of-roundness. Progressive increases in axial gap length due to cladding elongation and fuel stack shrinkage are modeled, as are deformations of fuel pellets and stainless steel support sleeves which bound plenum axial gaps in LWBR type blanket fuel rods. Zircaloy-4 primary and secondary thermal creep representations were developed from uniaxial creep testing of fuel rod tubing. Creep response to multi-axial loading is modeled with a variation of Hill's formulation for anisotropic materials. Coefficients accounting for anisotropic thermal creep in Zircaloy-4 tubes were developed from creep testing of externally pressurized tubes having fixed axial gaps in the range 2.5 cm to 5.7 cm and radial clearances over simulated fuel pellets ranging from zero to 0.089 mm. (orig./RW)

  13. Thermophysical instruments for non-destructive examination of tightness and internal gas pressure or irradiated power reactor fuel rods

    International Nuclear Information System (INIS)

    Pastoushin, V.V.; Novikov, A.Yu.; Bibilashvili, Yu.K.

    1998-01-01

    The developed thermophysical method and technical instruments for non-destructive leak-tightness and gas pressure inspection inside irradiated power reactor fuel rods and FAs under poolside and hot cell conditions are described. The method of gas pressure measuring based on the examination of parameters of thermal convection that aroused in gas volume of rod plenum by special technical instruments. The developed method and technique allows accurate value determination of not only one of the main critical rod parameters, namely total internal gas pressure, that forms rod mean life in the reactor core, but also the partial pressure of every main constituent of gaseous mixture inside irradiated fuel rod, that provides the feasibility of authentic and reliable leak-tightness detection. The described techniques were experimentally checked during the examination of all types power reactor fuel rods existing in Russia (WWER, BN, RBMK) and could form the basis for new technique development for non-destructive examination of PWR (and other) type rods and FAs having gas plenum filled with spring or another elements of design. (author)

  14. Experimental and CFD Studies of Coolant Flow Mixing within Scaled Models of the Upper and Lower Plenums of NGNP Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Anand, Nk [Texas A & M Univ., College Station, TX (United States)

    2016-03-30

    A 1/16th scaled VHTR experimental model was constructed and the preliminary test was performed in this study. To produce benchmark data for CFD validation in the future, the facility was first run at partial operation with five pipes being heated. PIV was performed to extract the vector velocity field for three adjacent naturally convective jets at statistically steady state. A small recirculation zone was found between the pipes, and the jets entered the merging zone at 3 cm from the pipe outlet but diverged as the flow approached the top of the test geometry. Turbulence analysis shows the turbulence intensity peaked at 41-45% as the jets mixed. A sensitivity analysis confirmed that 1000 frames were sufficient to measure statistically steady state. The results were then validated by extracting the flow rate from the PIV jet velocity profile, and comparing it with an analytic flow rate and ultrasonic flowmeter; all flow rates lie within the uncertainty of the other two methods for Tests 1 and 2. This test facility can be used for further analysis of naturally convective mixing, and eventually produce benchmark data for CFD validation for the VHTR during a PCC or DCC accident scenario. Next, a PTV study of 3000 images (1500 image pairs) were used to quantify the velocity field in the upper plenum. A sensitivity analysis confirmed that 1500 frames were sufficient to precisely estimate the flow. Subsequently, three (3, 9, and 15 cm) Y-lines from the pipe output were extracted to consider the output differences between 50 to 1500 frames. The average velocity field and standard deviation error that accrued in the three different tests were calculated to assess repeatability. The error was varied, from 1 to 14%, depending on Y-elevation. The error decreased as the flow moved farther from the output pipe. In addition, turbulent intensity was calculated and found to be high near the output. Reynolds stresses and turbulent intensity were used to validate the data by

  15. Experimental and CFD Studies of Coolant Flow Mixing within Scaled Models of the Upper and Lower Plenums of NGNP Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Hassan, Yassin; Anand, Nk

    2016-01-01

    A 1/16th scaled VHTR experimental model was constructed and the preliminary test was performed in this study. To produce benchmark data for CFD validation in the future, the facility was first run at partial operation with five pipes being heated. PIV was performed to extract the vector velocity field for three adjacent naturally convective jets at statistically steady state. A small recirculation zone was found between the pipes, and the jets entered the merging zone at 3 cm from the pipe outlet but diverged as the flow approached the top of the test geometry. Turbulence analysis shows the turbulence intensity peaked at 41-45% as the jets mixed. A sensitivity analysis confirmed that 1000 frames were sufficient to measure statistically steady state. The results were then validated by extracting the flow rate from the PIV jet velocity profile, and comparing it with an analytic flow rate and ultrasonic flowmeter; all flow rates lie within the uncertainty of the other two methods for Tests 1 and 2. This test facility can be used for further analysis of naturally convective mixing, and eventually produce benchmark data for CFD validation for the VHTR during a PCC or DCC accident scenario. Next, a PTV study of 3000 images (1500 image pairs) were used to quantify the velocity field in the upper plenum. A sensitivity analysis confirmed that 1500 frames were sufficient to precisely estimate the flow. Subsequently, three (3, 9, and 15 cm) Y-lines from the pipe output were extracted to consider the output differences between 50 to 1500 frames. The average velocity field and standard deviation error that accrued in the three different tests were calculated to assess repeatability. The error was varied, from 1 to 14%, depending on Y-elevation. The error decreased as the flow moved farther from the output pipe. In addition, turbulent intensity was calculated and found to be high near the output. Reynolds stresses and turbulent intensity were used to validate the data by

  16. Ultrafast gas switching experiments

    International Nuclear Information System (INIS)

    Frost, C.A.; Martin, T.H.; Patterson, P.E.; Rinehart, L.F.; Rohwein, G.J.; Roose, L.D.; Aurand, J.F.; Buttram, M.T.

    1993-01-01

    We describe recent experiments which studied the physics of ultrafast gas breakdown under the extreme overvoltages which occur when a high pressure gas switch is pulse charged to hundreds of kV in 1 ns or less. The highly overvolted peaking gaps produce powerful electromagnetic pulses with risetimes Khz at > 100 kV/m E field

  17. Mitigation of thermal transients by tube bundle inlet plenum design

    International Nuclear Information System (INIS)

    Oras, J.J.; Kasza, K.E.

    1984-06-01

    A multiphase program aimed at investigating the importance of thermal buoyancy to LMFBR steam-generator and heat-exchanger thermal hydraulics under low-flow transient conditions is being conducted in the Argonne Mixing Components Test Facility (MCTF) on a 60 0 sector shell-side flow model of the Westinghouse straight-tube steam generator being developed under the US/DOE large-component development program. A series of shell-side constant-flow thermal-downramp transient tests have been conducted focusing on the phenomenon of thermal-buoyancy-induced-flow channeling. In addition, it was discovered that a shell-inlet flow-distribution plenum can play a significant role in mitigating the severity of a thermal transient entering a steam generator or heat exchanger

  18. Thermophysical, hydrodynamic and mechanical aspects of molten core relocation to lower plenum

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y; Huh, Chang Wook [Seoul National University, Seoul (Korea, Republic of)

    1998-12-31

    This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regard to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard Power Plant (KSNPP) reactor. The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective. 10 refs., 1 fig., 1 tab. (Author)

  19. Thermophysical, hydrodynamic and mechanical aspects of molten core relocation to lower plenum

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Kune Y.; Huh, Chang Wook [Seoul National University, Seoul (Korea, Republic of)

    1997-12-31

    This paper presents the current state of knowledge on molten material relocation into the lower plenum. Consequences of movement of material to the lower head are considered with regard to the potential for reactor pressure vessel failure from both thermal hydraulic and mechanical standpoints. The models are applied to evaluating various in-vessel retention strategies for the Korean Standard Power Plant (KSNPP) reactor. The results are summarized in terms of thermal response of the reactor vessel from the very relevant severe accident management perspective. 10 refs., 1 fig., 1 tab. (Author)

  20. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, December 1, 1975--February 29, 1976

    International Nuclear Information System (INIS)

    Todreas, N.E.; Golay, M.W.

    1976-01-01

    Progress is summarized in the following task areas: assessment of available data, experimental water mixing investigations, analytic model development, and analytical and experimental investigation of velocity and temperature fields in outlet plenum flow mixing

  1. Prediction of corium debris characteristics in lower plenum of a nordic BWR in different accident scenarios using MELCOR code - 15367

    International Nuclear Information System (INIS)

    Phung, V.A.; Galushin, S.; Raub, S.; Goronovski, A.; Villanueva, W.; Koeoep, K; Grishchenko, D.; Kudinov, P.

    2015-01-01

    Severe accident management strategy in Nordic boiling water reactors (BWRs) relies on ex-vessel core debris coolability. The mode of corium melt release from the vessel determines conditions for ex-vessel accident progression and threats to containment integrity, e.g., formation of a non-coolable debris bed and possibility of energetic steam explosion. In-vessel core degradation and relocation is an important stage which determines characteristics of corium debris in the vessel lower plenum, such as mass, composition, thermal properties, timing of relocation, and decay heat. These properties affect debris reheating and remelting, melt interactions with the vessel structures, and possibly vessel failure and melt ejection mode. Core degradation and relocation is contingent upon the accident scenario parameters such as recovery time and capacity of safety systems. The goal of this work is to obtain a better understanding of the impact of the accident scenarios and timing of the events on core relocation phenomena and resulting properties of the debris bed in the vessel lower plenum of Nordic BWRs. In this study, severe accidents in a Nordic BWR reference plant are initiated by a station black out event, which is the main contributor to core damage frequency of the reactor. The work focuses on identifying ranges of debris bed characteristics in the lower plenum as functions of the accident scenario with different recovery timing and capacity of safety systems. The severe accident analysis code MELCOR coupled with GA-IDPSA is used in this work. GA-IDPSA is a Genetic Algorithm-based Integrated Deterministic Probabilistic Safety Analysis tool, which has been developed to search uncertain input parameter space. The search is guided by different target functions. Scenario grouping and clustering approach is applied in order to estimate the ranges of debris characteristics and identify scenario regions of core relocation that can lead to significantly different debris bed

  2. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Progress report, March 1, 1977--May 31, 1977

    International Nuclear Information System (INIS)

    Todreas, N.E.; Golay, M.W.; Wolf, L.

    1977-01-01

    Progress is summarized in the following tasks: (1) bundle flow studies (wrapped and bare rods); (2) subchannel flow studies (bare rods); (3) LMFBR outlet plenum flow mixing; and (4) theoretical determination of local temperature fields in LMFBR fuel rod bundles

  3. Inhalational anaesthesia with low fresh gas flow

    Directory of Open Access Journals (Sweden)

    Christian Hönemann

    2013-01-01

    Full Text Available During the inhalation of anaesthesia use of low fresh gas flow (0.35-1 L/min has some important advantages. There are three areas of benefit: pulmonary - anaesthesia with low fresh gas flow improves the dynamics of inhaled anaesthesia gas, increases mucociliary clearance, maintains body temperature and reduces water loss. Economic - reduction of anaesthesia gas consumption resulting in significant savings of > 75% and Ecological - reduction in nitrous oxide consumption, which is an important ozone-depleting and heat-trapping greenhouse gas that is emitted. Nevertheless, anaesthesia with high fresh gas flows of 2-6 L/min is still performed, a technique in which rebreathing is practically negligible. This special article describes the clinical use of conventional plenum vaporizers, connected to the fresh gas supply to easily perform low (1 L/min, minimal (0.5 L/min or metabolic flow anaesthesia (0.35 L/min with conventional Primus Draeger® anaesthesia machines in routine clinical practice.

  4. A "Greenhouse Gas" Experiment for the Undergraduate Laboratory

    Science.gov (United States)

    Gomez, Elaine; Paul, Melissa; Como, Charles; Barat, Robert

    2014-01-01

    This experiment and analysis offer an effective experience in greenhouse gas reduction. Ammoniated water is flowed counter-current to a simulated flue gas of air and CO2 in a packed column. The gaseous CO2 concentrations are measured with an on-line, non- dispersive, infrared analyzer. Column operating parameters include total gas flux, dissolved…

  5. Modeling for evaluation of debris coolability in lower plenum of reactor pressure vessel

    International Nuclear Information System (INIS)

    Maruyama, Yu; Moriyama, Kiyofumi; Nakamura, Hideo; Hirano, Masashi

    2003-01-01

    Effectiveness of debris cooling by water that fills a gap between the debris and the lower head wall was estimated through steady calculations in reactor scale. In those calculations, the maximum coolable debris depth was assessed as a function of gap width with combination of correlations for critical heat flux and turbulent natural convection of a volumetrically heated pool. The results indicated that the gap with a width of 1 to 2 mm was capable of cooling the debris under the conditions of the TMI-2 accident, and that a significantly larger gap width was needed to retain a larger amount of debris within the lower plenum. Transient models on gap growth and water penetration into the gap were developed and incorporated into CAMP code along with turbulent natural convection model developed by Yin, Nagano and Tsuji, categorized in low Reynolds number type two-equation model. The validation of the turbulent model was made with the UCLA experiment on natural convection of a volumetrically heated pool. It was confirmed that CAMP code predicted well the distribution of local heat transfer coefficients along the vessel inner surface. The gap cooling model was validated by analyzing the in-vessel debris coolability experiments at JAERI, where molten Al 2 O 3 was poured into a water-filled hemispherical vessel. The temperature history measured on the vessel outer surface was satisfactorily reproduced by CAMP code. (author)

  6. Experience with unconventional gas turbine fuels

    Energy Technology Data Exchange (ETDEWEB)

    Mukherjee, D K [ABB Power Generation Ltd., Baden (Switzerland)

    1997-12-31

    Low grade fuels such as Blast Furnace Gas, biomass, residual oil, coke, and coal - if used in conjunction with appropriate combustion, gasification, and clean-up processes and in combination with a gas turbine combined cycle -offer attractive and environmentally sound power generation. Recently, the Bao Shan Iron and Steel Company in Shanghai placed an order with Kawasaki Heavy Industries, Japan, to supply a combined-cycle power plant. The plant is to employ ABB`s GT 11N2 with a combustor modified to burn blast furnace gas. Recent tests in Shanghai and at Kawasaki Steel, Japan, have confirmed the burner design. The same basic combustor concept can also be used for the low BTU gas derived from airblown gasification processes. ABB is also participating in the API project: A refinery-residual gasification combined-cycle plant in Italy. The GT 13E2 gas turbine employees MBTU EV burners that have been successfully tested under full operating conditions. These burners can also handle the MBTU gas produced in oxygenblown coal gasification processes. ABB`s vast experience in burning blast furnace gas (21 plants built during the 1950s and 1960s), residuals, crude, and coal in various gas turbine applications is an important asset for building such power plants. This presentation discusses some of the experience gained in such plants. (orig.) 6 refs.

  7. Experience with unconventional gas turbine fuels

    Energy Technology Data Exchange (ETDEWEB)

    Mukherjee, D.K. [ABB Power Generation Ltd., Baden (Switzerland)

    1996-12-31

    Low grade fuels such as Blast Furnace Gas, biomass, residual oil, coke, and coal - if used in conjunction with appropriate combustion, gasification, and clean-up processes and in combination with a gas turbine combined cycle -offer attractive and environmentally sound power generation. Recently, the Bao Shan Iron and Steel Company in Shanghai placed an order with Kawasaki Heavy Industries, Japan, to supply a combined-cycle power plant. The plant is to employ ABB`s GT 11N2 with a combustor modified to burn blast furnace gas. Recent tests in Shanghai and at Kawasaki Steel, Japan, have confirmed the burner design. The same basic combustor concept can also be used for the low BTU gas derived from airblown gasification processes. ABB is also participating in the API project: A refinery-residual gasification combined-cycle plant in Italy. The GT 13E2 gas turbine employees MBTU EV burners that have been successfully tested under full operating conditions. These burners can also handle the MBTU gas produced in oxygenblown coal gasification processes. ABB`s vast experience in burning blast furnace gas (21 plants built during the 1950s and 1960s), residuals, crude, and coal in various gas turbine applications is an important asset for building such power plants. This presentation discusses some of the experience gained in such plants. (orig.) 6 refs.

  8. Development of Lower Plenum Molten Pool Module of Severe Accident Analysis Code in Korea

    International Nuclear Information System (INIS)

    Son, Donggun; Kim, Dong-Ha; Park, Rae-Jun; Bae, Jun-Ho; Shim, Suk-Ku; Marigomen, Ralph

    2014-01-01

    To simulate a severe accident progression of nuclear power plant and forecast reactor pressure vessel failure, we develop computational software called COMPASS (COre Meltdown Progression Accident Simulation Software) for whole physical phenomena inside the reactor pressure vessel from a core heat-up to a vessel failure. As a part of COMPASS project, in the first phase of COMPASS development (2011 - 2014), we focused on the molten pool behavior in the lower plenum, heat-up and ablation of reactor vessel wall. Input from the core module of COMPASS is relocated melt composition and mass in time. Molten pool behavior is described based on the lumped parameter model. Heat transfers in between oxidic, metallic molten pools, overlying water, steam and debris bed are considered in the present study. The models and correlations used in this study are appropriately selected by the physical conditions of severe accident progression. Interaction between molten pools and reactor vessel wall is also simulated based on the lumped parameter model. Heat transfers between oxidic pool, thin crust of oxidic pool and reactor vessel wall are considered and we solve simple energy balance equations for the crust thickness of oxidic pool and reactor vessel wall. As a result, we simulate a benchmark calculation for APR1400 nuclear power plant, with assumption of relocated mass from the core is constant in time such that 0.2ton/sec. We discuss about the molten pool behavior and wall ablation, to validate our models and correlations used in the COMPASS. Stand-alone SIMPLE program is developed as the lower plenum molten pool module for the COMPASS in-vessel severe accident analysis code. SIMPLE program formulates the mass and energy balance for water, steam, particulate debris bed, molten corium pools and oxidic crust from the first principle and uses models and correlations as the constitutive relations for the governing equations. Limited steam table and the material properties are provided

  9. TRIO a general computer code for reactor 3-D flows analysis. Application to a LMFBR hot plenum

    International Nuclear Information System (INIS)

    Magnaud, J.P.; Rouzaud, P.

    1985-09-01

    TRIO is a code developed at CEA to investigate general incompressible 2D and 3D viscous flows. Two calculations are presented: the lid driven cubic cavity at Re=400; steady state (velocity and temperature field) of a LMFBR hot plenum, carried out in order to prepare the calculation of a cold shock consecutive to a reactor scram. 8 refs., 26 figs.

  10. Measurement of two-phase flow at the core upper plenum interface under simulated reflood conditions

    International Nuclear Information System (INIS)

    Thomas, D.G.; Combs, S.K.; Bagwell, M.E.

    1980-01-01

    Objectives of the Instrument Development Loop program were to simulate flows at the core/upper plenum interface during the reflood phase of a LOCA and to develop instruments for measuring mass-flows at this interface. A tie plate drag body was developed and tested successfully, and the data obtained were shown to be equivalent to pressure drops. The tie-plate drag body gave useful measurements in pure downflow, and the drag/turbine combination correlates with mass flow for high upflow

  11. Two-branch Gas Experiments for Hot Gas Mixing of HTR-PM

    International Nuclear Information System (INIS)

    Zhou Yangping; Hao Pengefei; He Heng; Li Fu; Shi Lei

    2014-01-01

    A model experiment is proposed to investigate the hot gas mixing efficiency of HTR-PM reactor outlet. The test facility is introduced which is set at a scale of 1:2.5 comparing with the design of thermal mixing structure at HTR-PM reactor outlet. The test facility using air as its flow media includes inlet pipe system, electric heaters, main body of test facility, hot gas duct, exhaust pipe system and I&C system. Two-branch gas experiments are conducted on the test facility and the values of thermal-fluid parameters are collected and analyzed which include the temperature, pressure and velocity of the flow as well as the temperature of the tube wall. The analysis result shows the mixing efficiency is higher than the requirement of thermal mixing by steam generator even with conservative assumption which indicates that the design of hog gas mixing structure of HTR-PM fulfills the requirement for thermal mixing at two-branch working conditions. (author)

  12. Experimental system description for air-water CCFL tests of the 161-rod FLECHT-SEASET test vessel upper plenum

    International Nuclear Information System (INIS)

    Fogdall, S.P.; Anderson, J.L.

    1983-01-01

    A series of countercurrent flow limiting (CCFL) experiments has been performed by EG and G Idaho, Inc. in the Steam-Air-Water (SAW) test facility at the Idaho National Engineering Laboratory on behalf of the US Nuclear Regulatory Commission (NRC). Tests were performed in a mockup of the vessel for the 161-Rod Systems Effects Test (SET) facility of the FLECHT-SEASET program, conducted by the Westinghouse Electric Corporation. Westinghouse and the NRC will use the test results to provide a CCFL correlation to predict the flooding behavior in the upper plenum of the SET vessel. This paper presents a description of the experimental system and the test conduct, including data validation and uncertainty analysis. The test objectives centered on experimentally obtaining coefficients in the Wallis correlation for flooding with the specific vessel geometry. The test conditions and vessel configuration are described and the design of the test loop, instrumentation, and data acquisition are discussed. The establishment of a test point and the resultant data are described

  13. The Spanish experience - future developments in the gas industry

    International Nuclear Information System (INIS)

    Moraleda, P.

    1996-01-01

    Spanish experience is presented concerned it may be useful at the time of setting up a natural gas industry. The Spanish natural gas industry is of recent creation. Developing infrastructure and securing gas supplies have been major challenges. Challenges which, are also common for majority of the countries. The presentation is split into two blocks: the first one is on our experience in the establishment and consolidation of the market for natural gas in Spain. The second block deals with future developments aiming to strengthen the security of supply; and with the opportunities and threats the gas industry will face

  14. Creys-Malville nuclear plant. Simulation of the cold plenum thermal-hydraulics. 12 zone model presentation

    International Nuclear Information System (INIS)

    Faulot, J.P.

    1990-05-01

    The CRUSIFI code has been developed by SEPTEN (Engineering and Construction Division) with SICLE software during 1983-1985 in order to study the CREYS-MALVILLE dynamic behavior. At the time, the version was based on project data (version 2.3). It includes a 2 zones model for the cold plenum thermal-hydraulics, modelling which does not allow to reproduce accurately dissymetries apt to occur as well in usual operating (hydraulic dissymetries bound to one or many systems out of order), as during incidentally operating (hydraulic dissymetries bound to primary pump working back or thermal dissymetries after a transient on one or many secondary loops). Moreover, a 2 zones model cannot simulate axial temperature gradients which appear during double stratification phenomenon (upper and lower part of the plenum) produced by alternating thermal shock. A 12 zones model (4 sectors with 3 axial zones each) such as model developed by R$DD (Research and Development Division) allows to satisfy correctly these problems. This report is a specification of the chosen modelling. This model is now operational after qualifying with experimental transients on mockup and reactor. It is to-day connected with the EDF general operating code CRUSIFI (calibrating version 3.0). It could be easily integrated in a four loops plant modelling such as the CREYS-MALVILLE simulator in a four loops plant modelling such as the CREYS-MALVILLE simulator under construction at the present time by THOMSON

  15. Experiment data report for semiscale MOD-1 test S-01-3 (isothermal blowdown with core resistance simulator)

    International Nuclear Information System (INIS)

    Zender, S.N.

    1975-03-01

    Recorded test data are presented for Test S-01-3 of the semiscale Mod-1 isothermal blowdown test series. Test S-01-3 is one of several semiscale Mod-1 experiments which are counterparts of the planned Loss-of-Fluid Test (LOFT) nonnuclear experiments. System hardware is of the LOFT design, selected using volumetric scaling methods, and initial conditions duplicate those identified for the LOFT nonnuclear tests. Test S-01-3 employed an intact loop resistance that was low relative to that of the first test in the series (Test S-01-2) to establish the importance of intact loop resistance on system response during blowdown. An orificed structure was used in the pressure vessel to simulate the LOFT core simulator. The test was initiated at isothermal conditions of 2245 psig and 538 0 F by a simulated offset shear of the cold-leg broken loop piping. During system depressurization, coolant was injected into the lower plenum of the pressure vessel to provide data on the effects of emergency core cooling on system response. Additionally, to aid in determination of the effects of accumulator gas on pressure suppression system response, the nitrogen used to charge the accumulator systems for Test S-01-3 was allowed to vent into the lower plenum following depletion of the coolant. (U.S.)

  16. Experience in education and training of gas engineers in Russia

    International Nuclear Information System (INIS)

    Basniev, K.; Vladimirov, A.

    1997-01-01

    Experience gained in training and retraining of engineers for gas industry is considered in the report. The report contains the material on modern state of higher technical education in Russia in view of the reforms taking place in this country. The report deals with questions concerning the experience gained in a specialized training of gas engineers at higher educational establishments of Russia including training of specialists for foreign countries. Conditions under which retraining of engineers involved in gas industry takes place are presented in the report. The report is based mainly on the experience gained by the Russian leading higher educational establishment of oil and gas profile, that is the State Gubkin Oil and Gas Academy. (au)

  17. Heavy ion source support gas mixing experiments

    International Nuclear Information System (INIS)

    Hudson, E.D.; Mallory, M.L.

    1977-01-01

    Experiments on mixing an easily ionized support gas with the primary ion source gas have produced large beam enhancements for high charge state light ions (masses less than or equal to 20). In the Oak Ridge Isochronous Cyclotron (ORIC), the beam increase has been a factor of 5 or greater, depending on ion species and charge state. Approximately 0.1 cc/min of the easily ionized support gas (argon, krypton, or xenon) is supplied to the ion source through a separate gas line and the primary gas flow is reduced by approximately 30 percent. The proposed mechanism for increased intensity is as follows: The heavier support gas ionizes readily to a higher charge state, providing increased cathode heating. The increased heating permits a reduction in primary gas flow (lower pressure) and the subsequent beam increase

  18. Ideal Gas Laws: Experiments for General Chemistry

    Science.gov (United States)

    Deal, Walter J.

    1975-01-01

    Describes a series of experiments designed to verify the various relationships implicit in the ideal gas equation and shows that the success of the Graham's law effusion experiments can be explained by elementary hydrodynamics. (GS)

  19. Analysis of turbulent natural convection heat transfer in a lower plenum during external cooling using the COSMO code

    Energy Technology Data Exchange (ETDEWEB)

    Noguchi, H. [Nuclear Power Engineering Corp., Tokyo (Japan); Sawatari, Y.; Imada, T. [Fuji Research Institute Corporation, Tokyo (Japan)

    2000-11-01

    The behavior of a large volumetrically heated melt pool is important to evaluate the feasibility of in-vessel retention by external flooding as an accident management. The COSMO (Coolability Simulation of Molten corium during severe accident) code has been developed at NUPEC to simulate turbulent natural convection heat transfer with internal heat source. The COSMO code solves thermal hydraulic conservation equations with turbulent model and can simulate melting and solidification process. The standard k-{epsilon} model has a limitation to describe the turbulent natural convection in the very high Rayleigh number condition (10{sup 16}-10{sup 17}) assumed to occur in a lower plenum of RPV during a severe accident. This limitation results from the assumption of an analogy of momentum and energy transfer phenomena in the standard model. In this paper the modified turbulent model in which the turbulent number is treated, as a function of the flux Richardson number derived from the experiment, has been incorporated and verified by using the BALI experiments. It was found that the prediction of averaged Nusselt number became better than that of the standard model. In order to extend the COSMO code to the actual scale analysis under the external flooding conditions, more realistic boundary condition derived from the experiments should be treated. In this work the CHF correlation from ULPU experiment or the heat transfer coefficient correlation from CYBL experiment have been applied. The preliminary analysis of an actual scale analysis has been carried out under the condition of the TMI-2 accident. (author)

  20. Analysis of turbulent natural convection heat transfer in a lower plenum during external cooling using the COSMO code

    International Nuclear Information System (INIS)

    Noguchi, H.; Sawatari, Y.; Imada, T.

    2000-01-01

    The behavior of a large volumetrically heated melt pool is important to evaluate the feasibility of in-vessel retention by external flooding as an accident management. The COSMO (Coolability Simulation of Molten corium during severe accident) code has been developed at NUPEC to simulate turbulent natural convection heat transfer with internal heat source. The COSMO code solves thermal hydraulic conservation equations with turbulent model and can simulate melting and solidification process. The standard k-ε model has a limitation to describe the turbulent natural convection in the very high Rayleigh number condition (10 16 -10 17 ) assumed to occur in a lower plenum of RPV during a severe accident. This limitation results from the assumption of an analogy of momentum and energy transfer phenomena in the standard model. In this paper the modified turbulent model in which the turbulent number is treated, as a function of the flux Richardson number derived from the experiment, has been incorporated and verified by using the BALI experiments. It was found that the prediction of averaged Nusselt number became better than that of the standard model. In order to extend the COSMO code to the actual scale analysis under the external flooding conditions, more realistic boundary condition derived from the experiments should be treated. In this work the CHF correlation from ULPU experiment or the heat transfer coefficient correlation from CYBL experiment have been applied. The preliminary analysis of an actual scale analysis has been carried out under the condition of the TMI-2 accident. (author)

  1. Acoustic sensor for in-pile fuel rod fission gas release measurement

    International Nuclear Information System (INIS)

    Fourmentel, D.; Villard, J. F.; Ferrandis, J. Y.; Augereau, F.; Rosenkrantz, E.; Dierckx, M.

    2009-01-01

    We have developed a specific acoustic sensor to improve the knowledge of fission gas release in Pressurized Water Reactor (PWR) fuel rods when irradiated in materials testing reactors. In order to perform experimental programs related to the study of the fission gas release kinetics, the CEA (French Nuclear Energy Commission) acquired the ability to equip a pre-irradiated PWR fuel rod with three sensors, allowing the simultaneous on-line measurements of the following parameters: - fuel temperature with a centre-line thermocouple type C, - internal pressure with a specific counter-pressure sensor, - fraction of fission gas released in the fuel rod with an innovative acoustic sensor. The third detector is the subject of this paper. This original acoustic sensor has been designed to measure the molar mass and pressure of the gas contained in the fuel rod plenum. For in-pile instrumentation, the fraction of fission gas, such as Krypton and Xenon, in Helium, can be deduced online from this measurement. The principle of this acoustical sensor is the following: a piezoelectric transducer generates acoustic waves in a cavity connected to the fuel rod plenum. The acoustic waves are propagated and reflected in this cavity and then detected by the transducer. The data processing of the signal gives the velocity of the acoustic waves and their amplitude, which can be related respectively to the molar mass and to the pressure of the gas. The piezoelectric material of this sensor has been qualified in nuclear conditions (gamma and neutron radiations). The complete sensor has also been specifically designed to be implemented in materials testing reactors conditions. For this purpose some technical points have been studied in details: - fixing of the piezoelectric sample in a reliable way with a suitable signal transmission, - size of the gas cavity to avoid any perturbation of the acoustic waves, - miniaturization of the sensor because of narrow in-pile experimental devices

  2. FIX-II. Loca-blowdown heat transfer and pump trip experiments. Summary report of phase 1: Design of experiments

    International Nuclear Information System (INIS)

    Waaranperae, Y.; Nilsson, L.; Gustafsson, P.Aa.; Jonsson, N.O.

    1979-06-01

    FIX-II is a loss of coolant blowdown heat transfer experiment, performed under contract for The Swedish Nuclear Power Inspectorate, SKI. The purpose of the experiments is to provide measurements from simulations of a pipe rupture on an external recirculation line in a Swedish BWR. Pump trips in BWRs with internal recirculation pumps will also be simulated. The existing FIX-loop at the Thermal Engineering Laboratory of Studsvik Energiteknik AB will be modified and used for the experiments. Components are included to simulate the steam dome, downcomer, two recirculation lines with one pump each, lower plenum, core (36-rod full length bundle), control rod guide tubes, core bypass, upper plenum and steam separators. The results of the first phase of the project are reported here. The following tasks are included in Phase 1: reactor reference analysis, scaling calculations of the FIX loop, development of fuel rod simulators, design of test section and test loop layout and proposal for test program. Further details of the work and results obtained for the different sub-projects are published in a number ofdetailed reports. (author)

  3. Gas box control system for Tandem Mirror Experiment-Upgrade

    International Nuclear Information System (INIS)

    Bell, H.H. Jr.; Hunt, A.L.; Clower, C.A. Jr.

    1983-01-01

    The Tandem Mirror Experiment-Upgrade (TMX-U) uses several methods to feed gas (usually deuterium) at different energies into the plasma region of the machine. One is an arrangement of eight high-speed piezo-electric valves mounted on special manifolds (gas box) that feed cold gas directly to the plasma. This paper describes the electronic valve control and data acquisition portions of the gas box, which are controlled by a desk-top computer. Various flow profiles have been developed and stored in the control computer for ready access by the operator. The system uses two modes of operation, one that exercises and characterizes the valves and one that operates the valves with the rest of the experiment. Both the valve control signals and the pressure transducers data are recorded on the diagnostics computer so that they are available for experiment analysis

  4. New piezo driven gas inlet valve for fusion experiments

    International Nuclear Information System (INIS)

    Usselmann, E.; Hemmerich, J.L.; How, J.; Holland, D.; Orchard, J.; Winkel, T.; Schargitz, U.; Pocheim, N.

    1989-01-01

    The gas inlet valves used at the JET experiment are described and their performances are discussed. A new gas-valve development suitable to replace the existing valves at JET and for future use in large fusion experiments is presented. The new valve is equipped with a piezo-electric translator and has a dosing range of 0-800 mbarls -1 for D 2 . The operating mode of the valve is fail-safe closed with a leak-rate of ≤ 10 -9 mbarls -1 . The design, the test results and throughput values in dependence of filling pressure and control voltage are presented and experiences with the prototype valve as a new gas inlet valve for the JET operation are described

  5. A fundamental study on sodium-water reaction in the double pool LMFBR, (3)

    International Nuclear Information System (INIS)

    Uotani, Masaki; Kumagai, Hiromichi; Nishi, Yoshihisa; Yoshida, Kazuo

    1989-01-01

    The double pool LMFBR is an innovative reactor that Central Research Institute of Electric Power Industry proposed for the purpose of reducing the construction cost of FBRs, and it is characterized by immersing steam generators in the annular plenum formed between the primary vessel and the outer secondary vessel. Therefore, it is expected that the pressure behavior at the time of sodium-water reaction due to the breaking of heating tubes is largely different from the case of steam generators of conventional FBRs. In order to ensure the soundness of the primary vessel that containes the reactor core, it is necessary to sufficiently grasp the pressure behavior in the plenum, and this basic experiment and analysis are related to the pressure behavior due to piston motion that arises in the initial period of quasi-steady pressure. About 1/10 scale annular plenum was used, and the generation of reaction product gas was simulated by the release of nitrogen. When gas was released in the plenum, the highest pressure rise occurred in the initial period of release, and thereafter, periodic variation arose. The pressure waveform and the value of pressure rise as the results of the model analysis agreed well with the measured results. (K.I.)

  6. Rarefied gas electro jet (RGEJ) micro-thruster for space propulsion

    Science.gov (United States)

    Blanco, Ariel; Roy, Subrata

    2017-11-01

    This article numerically investigates a micro-thruster for small satellites which utilizes plasma actuators to heat and accelerate the flow in a micro-channel with rarefied gas in the slip flow regime. The inlet plenum condition is considered at 1 Torr with flow discharging to near vacuum conditions (consumption and the thrust effectiveness of the thruster are predicted based on these results. The ionized gas is modelled using local mean energy approximation. An electrically induced body force and a thermal heating source are calculated based on the space separated charge distribution and the ion Joule heating, respectively. The rarefied gas flow with these electric force and heating source is modelled using density-based compressible flow equations with slip flow boundary conditions. The results show that a significant improvement of specific impulse can be achieved over highly optimized cold gas thrusters using the same propellant.

  7. Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

    International Nuclear Information System (INIS)

    Todreas, N.E.; Cheng, S.K.; Basehore, K.

    1984-08-01

    This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized. Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified

  8. IFPE/RISOE-II, Fuel Performance Data from Transient Fission Gas Release

    International Nuclear Information System (INIS)

    Turnbull, J.A.

    1995-01-01

    Description: The RISO National Laboratory in Denmark have carried out three irradiation programs of slow ramp and hold tests, so called 'bump tests' to investigate fission gas release and fuel microstructural changes. The second project took place between 1982 and 1986 and was called 'The RISO Transient Fission Gas Project'. The fuel used in the project was from: IFA-161 irradiated in the Halden BWR (27 to 42 MWd/kgUO 2 ) and GE BWR fuel irradiated in the Millstone 1 reactor 14 to 29 MWd/kgUO 2 . Using the re-fabrication technique, it was possible to back fill the test segment with a choice of gas and gas pressure and to measure the time dependence of fission gas release by continuous monitoring of the plenum pressure. The short length of the test segment was an advantage because, depending on where along the original rod the section was taken, burnup could be chosen variable, and during the test the fuel experienced a single power

  9. Critical heat flux of water in vertical tubes with an upper plenum and a closed bottom

    International Nuclear Information System (INIS)

    Kim, Hong Chae; Baek, Won Pil; Chang, Soon Heung

    2000-01-01

    An experimental study is conducted for vertical round tubes with an upper plenum and a closed bottom to investigate CHF behavior and CHF onset location under the counter-current condition. The measured CHF values are well predicted by general Wallis type flooding correlations. A 1-D steady state analytical flooding model for thermosyphon by El-Genk and Saber was assessed with the data and the liquid film thickness at the liquid entrance was calculated. The CHF onset position becomes different with L/D and D, and liquid entrance geometry affects only CHF values not CHF onset positions

  10. Flow design and simulation of a gas compression system for hydrogen fusion energy production

    Energy Technology Data Exchange (ETDEWEB)

    Avital, E J; Salvatore, E [School of Engineering and Materials Science, Queen Mary University of London, Mile End Rd London E1 4NS (United Kingdom); Munjiza, A [Civil Engineering, University of Split, Livanjska 2100 Split (Croatia); Suponitsky, V; Plant, D; Laberge, M, E-mail: e.avital@qmul.ac.uk [General Fusion Inc.,108-3680 Bonneville Place, Burnaby, BC V3N 4T5 (Canada)

    2017-08-15

    An innovative gas compression system is proposed and computationally researched to achieve a short time response as needed in engineering applications such as hydrogen fusion energy reactors and high speed hammers. The system consists of a reservoir containing high pressure gas connected to a straight tube which in turn is connected to a spherical duct, where at the sphere’s centre plasma resides in the case of a fusion reactor. Diaphragm located inside the straight tube separates the reservoir’s high pressure gas from the rest of the plenum. Once the diaphragm is breached the high pressure gas enters the plenum to drive pistons located on the inner wall of the spherical duct that will eventually end compressing the plasma. Quasi-1D and axisymmetric flow formulations are used to design and analyse the flow dynamics. A spike is designed for the interface between the straight tube and the spherical duct to provide a smooth geometry transition for the flow. Flow simulations show high supersonic flow hitting the end of the spherical duct, generating a return shock wave propagating upstream and raising the pressure above the reservoir pressure as in the hammer wave problem, potentially giving temporary pressure boost to the pistons. Good agreement is revealed between the two flow formulations pointing to the usefulness of the quasi-1D formulation as a rapid solver. Nevertheless, a mild time delay in the axisymmetric flow simulation occurred due to moderate two-dimensionality effects. The compression system is settled down in a few milliseconds for a spherical duct of 0.8 m diameter using Helium gas and a uniform duct cross-section area. Various system geometries are analysed using instantaneous and time history flow plots. (paper)

  11. Flow design and simulation of a gas compression system for hydrogen fusion energy production

    Science.gov (United States)

    Avital, E. J.; Salvatore, E.; Munjiza, A.; Suponitsky, V.; Plant, D.; Laberge, M.

    2017-08-01

    An innovative gas compression system is proposed and computationally researched to achieve a short time response as needed in engineering applications such as hydrogen fusion energy reactors and high speed hammers. The system consists of a reservoir containing high pressure gas connected to a straight tube which in turn is connected to a spherical duct, where at the sphere’s centre plasma resides in the case of a fusion reactor. Diaphragm located inside the straight tube separates the reservoir’s high pressure gas from the rest of the plenum. Once the diaphragm is breached the high pressure gas enters the plenum to drive pistons located on the inner wall of the spherical duct that will eventually end compressing the plasma. Quasi-1D and axisymmetric flow formulations are used to design and analyse the flow dynamics. A spike is designed for the interface between the straight tube and the spherical duct to provide a smooth geometry transition for the flow. Flow simulations show high supersonic flow hitting the end of the spherical duct, generating a return shock wave propagating upstream and raising the pressure above the reservoir pressure as in the hammer wave problem, potentially giving temporary pressure boost to the pistons. Good agreement is revealed between the two flow formulations pointing to the usefulness of the quasi-1D formulation as a rapid solver. Nevertheless, a mild time delay in the axisymmetric flow simulation occurred due to moderate two-dimensionality effects. The compression system is settled down in a few milliseconds for a spherical duct of 0.8 m diameter using Helium gas and a uniform duct cross-section area. Various system geometries are analysed using instantaneous and time history flow plots.

  12. Cascading Tesla Oscillating Flow Diode for Stirling Engine Gas Bearings

    Science.gov (United States)

    Dyson, Rodger

    2012-01-01

    Replacing the mechanical check-valve in a Stirling engine with a micromachined, non-moving-part flow diode eliminates moving parts and reduces the risk of microparticle clogging. At very small scales, helium gas has sufficient mass momentum that it can act as a flow controller in a similar way as a transistor can redirect electrical signals with a smaller bias signal. The innovation here forces helium gas to flow in predominantly one direction by offering a clear, straight-path microchannel in one direction of flow, but then through a sophisticated geometry, the reversed flow is forced through a tortuous path. This redirection is achieved by using microfluid channel flow to force the much larger main flow into this tortuous path. While microdiodes have been developed in the past, this innovation cascades Tesla diodes to create a much higher pressure in the gas bearing supply plenum. In addition, the special shape of the leaves captures loose particles that would otherwise clog the microchannel of the gas bearing pads.

  13. RANS simulation of the thermal mixing in HTTF LP during normal operation conditions – High Temperature Test Facility at Oregon State University

    International Nuclear Information System (INIS)

    Gradecka, Malwina J.; Woods, Brian

    2014-01-01

    Since High Temperature Gas-cooled Reactors are being considered as the most promising design of upcoming IV Gen reactors, key research areas were identified to address safety aspects of this design. A number of simulations and experiments need to be conducted in this field. In this paper, thermal-hydraulics aspects of coolant flow through Lower Plenum (LP) of HTGR were considered, specifically flow characteristics to identify the risk of temperature stratification in LP and hot spotting on LP floor. Local temperature gradients can cause material degradation. As the power profile is non-uniform across the core, jets of coolant exit the core region at different temperatures and enter the LP impinging on LP floor causing hot spots at LP structure and temperature stratification. To address those issues numerical simulation and an experiment are being developed. The numerical simulation provides coolant flow velocity and temperature fields. The purpose of this study is to investigate the mixing phenomenon in the LP due to risk of the hot streaking and thermal stratification phenomena during normal operation of HTTF. The following aspect are being examined: identification of gas flow behavior in lower plenum of HTTF based on CFD simulations, identification of hot streaking issue in the HTTF lower plenum using CFD tools, and computational investigation of gas mixing efficiency. This paper includes a description of experimental setup of HTTF, guidance for LP CFD modeling, and the results and analysis of CFD simulation. (author)

  14. Development and verification of the LIFE-GCFR computer code for predicting gas-cooled fast-reactor fuel-rod performance

    International Nuclear Information System (INIS)

    Hsieh, T.C.; Billone, M.C.; Rest, J.

    1982-03-01

    The fuel-pin modeling code LIFE-GCFR has been developed to predict the thermal, mechanical, and fission-gas behavior of a Gas-Cooled Fast Reactor (GCFR) fuel rod under normal operating conditions. It consists of three major components: thermal, mechanical, and fission-gas analysis. The thermal analysis includes calculations of coolant, cladding, and fuel temperature; fuel densification; pore migration; fuel grain growth; and plenum pressure. Fuel mechanical analysis includes thermal expansion, elasticity, creep, fission-product swelling, hot pressing, cracking, and crack healing of fuel; and thermal expansion, elasticity, creep, and irradiation-induced swelling of cladding. Fission-gas analysis simultaneously treats all major mechanisms thought to influence fission-gas behavior, which include bubble nucleation, resolution, diffusion, migration, and coalescence; temperature and temperature gradients; and fission-gas interaction with structural defects

  15. Studies of flow stratification in the hot plenum of an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Jones, P; Hickmott, S [Central Electricity Generating Board, Berkeley Nuclear Laboratories, Berkeley, Gloucestershire (United Kingdom)

    1983-07-01

    The paper reviews work at Berkeley Nuclear Laboratories on the extent and effects of buoyancy in the hot plenum of an LMFBR. It summarizes the experimental, theoretical and numerical work has has been conducted to aid the understanding of the complex transient flows which occur following a reactor trip. The experimental work has been conducted in small-scale idealised geometries which isolate the essential features of the reactor flows and is not intended to provide detailed design data. An integral theory has been devised to describe the thermal hydraulics of negatively-buoyant jets. The predictions are shown to be in good agreement with the experimental results and emphasize the need to correctly represent the inlet velocity and temperature profiles. Some preliminary calculations with a transient, two-dimensional, finite-element code are compared with the experimental results. These calculations reproduce the overall features of the flows but not the details of the stratified interface. The development of turbulence models for stratified flows is seen as a fruitful area for further research. (author)

  16. Experiment data report for Semiscale Mod-1 Test S-05-4 (alternate ECC injection test)

    International Nuclear Information System (INIS)

    Collins, B.L.; Feldman, E.M.

    1977-03-01

    Recorded test data are presented for Test S-05-4 of the Semiscale Mod-1 alternate emergency core coolant injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-4 was conducted from initial conditions of 2266 psia and 543 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg of each loop and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. The upper plenum coolant injection was scaled according to the heat stored in the metal mass of the upper plenum

  17. Ruthenium removing device

    International Nuclear Information System (INIS)

    Kitamura, Masafumi; Shirado, Katsuyuki.

    1990-01-01

    A processing gas supply system and a NO x supply system for supplying NO x to be mixed with processed gases are connected to the gas plenum in the lower portion of reaction vessel. Further, a cleaning station is disposed above the gas plenum for introducing a mixed gas stream from the gas plenum into a liquid detergent thereby trapping NO x and ruthenium reduction products into the liquid detergent. Volatile ruthenium contained in the processed gases is reduced into ruthenium reduction products and formed as mists. They are trapped in the cleaning liquid and the remaining gases are discharged out of the liquid detergent to the outside of the reaction vessel. Accordingly, solid radioactive wastes are not formed and the decontaminating efficiency for volatile ruthenium can be improved. (T.M.)

  18. Transient Fuel Behavior and Failure Condition in the CABRI-2 Experiments

    International Nuclear Information System (INIS)

    Sato, Ikken; Lemoine, Francette; Struwe, Dankward

    2004-01-01

    In the CABRI-2 program, 12 tests were performed under various transient conditions covering a wide range of accident scenarios using two types of preirradiated fast breeder reactor (FBR) fuel pins with different smear densities and burnups. For each fuel, a nonfailure-transient test was performed, and it provided basic information such as fuel thermal condition, fuel swelling, and gas release. From the failure tests, information on failure mode, failure time, and axial location was obtained. Based on this information, failure conditions such as fuel enthalpy and cladding temperature were evaluated. These failure conditions were compared with the CABRI-1 tests in which different fuels as well as different transient conditions were used. This comparison, together with supporting information available from existing in-pile and out-of-pile experiments, allowed an effective understanding on failure mechanisms depending on fuel and transient conditions. It is concluded that pellet-cladding mechanical interaction (PCMI) due to fuel thermal expansion and fission-gas-induced swelling is playing an important role on mechanical clad loading especially with high smear density and low fuel-heating-rate conditions. At very high heating-rate conditions, there is no sufficient time to allow significant fuel swelling, so that cavity pressurization with fuel melting becomes the likely failure mechanism. Fuel smear density and fission-gas retention have a strong impact both on PCMI and cavity pressurization. Furthermore, pin failure is strongly dependent on cladding temperature, which plays an important role in the axial failure location. With the low smear-density fuel, considerable PCMI mitigation is possible leading to a high failure threshold as well as in-pin molten-fuel relocation along the central hole. However, even with the low smear density fuel, PCMI failure could take place with an elevated cladding-temperature condition. On the other hand, in case of a sufficiently long

  19. Estimate of LOCA-FI plenum pressure uncertainty for a five-ring RELAP5 production reactor model

    International Nuclear Information System (INIS)

    Griggs, D.P.

    1993-03-01

    The RELAP5/MOD2.5 code (RELAP5) is used to perform best-estimate analyses of certain postulated Design Basis Accidents (DBAs) in SRS production reactors. Currently, the most limiting DBA in terms of reactor power level is an instantaneous double-ended guillotine break (DEGB) loss of coolant accident (LOCA). A six-loop RELAP5 K Reactor model is used to analyze the reactor system behavior dozing the Flow Instability (FI) phase of the LOCA, which comprises only the first 5 seconds following the DEGB. The RELAP5 K Reactor model includes tank and plenum nodalizations having five radial rings and six azimuthal sectors. The reactor system analysis provides time-dependent plenum and tank bottom pressures for use as boundary conditions in the FLOWTRAN code, which models a single fuel assembly in detail. RELAP5 also performs the system analysis for the latter phase of the LOCA, denoted the Emergency Cooling System (ECS) phase. Results from the RELAP analysis are used to provide boundary conditions to the FLOWTRAN-TF code, which is an advanced two-phase version of FLOWTRAN. The RELAP5 K Reactor model has been tested for LOCA-FI and Loss-of-Pumping Accident analyses and the results compared with equivalent analyses performed with the TRAC-PF1/MOD1 code (TRAC). An equivalent RELAP5 six-loop, five-ring, six-sector L Reactor model has been benchmarked against qualified single-phase system data from the 1989 L-Area In-Reactor Test Program. The RELAP5 K and L Reactor models have also been subjected to an independent Quality Assurance verification

  20. Issues of Exercising the Right to Defence amid the Explanations of the Plenum of the Supreme Court of the Russian Federation

    Directory of Open Access Journals (Sweden)

    Oksana A. Voltornist

    2016-04-01

    Full Text Available The article analyzes the explanations of the Plenum of the Supreme Court No. 29 dated June 30, 2015 “On application of laws by the courts ensuring the right to defense in criminal proceedings”. The author details the applied aspects of certain provisions of the aforementioned document within the criminal procedure legislation and estimates their significance for the judicial and investigative practice

  1. Three-dimensional calculation of the flow in the cold plenum of the Fast Breeder Reactor 1500 MW

    International Nuclear Information System (INIS)

    Chabard, J.P.; Daubert, O.; Gregoire, J.P.; Hemmerich, P.

    1987-01-01

    To solve thermalhydraulics problems which are rising for example on the various parts of nuclear reactors, several departments of the Direction des Etudes et Recherches are developing the N3S code, three-dimensional code using the finite element method. First, this paper presents the basic equations (Navies-Stokes with turbulence modelling and coupled with the thermal equation) and well suited algorithms to solve them. The industrial adequacy of the code is clearly demonstrated through the application to the computation of the flow in the cold plenum of the Fast Breeder Reactor 1500 MW on a mesh of about 20000 velocity nodes [fr

  2. Numerical modeling of the plasma ring acceleration experiment

    International Nuclear Information System (INIS)

    Eddleman, J.L.; Hammer, J.H.; Hartman, C.W.

    1987-01-01

    Modeling of the LLNL RACE experiment and its many applications has necessitated the development and use of a wide array of computational tools. The two-dimensional MHD code, HAM, has been used to model the formation of a compact torus plasma ring in a magnetized coaxial gun and its subsequent acceleration by an additional applied toroidal field. Features included in the 2-D calculations are self-consistent models for (1) the time-dependent poloidal field produced by a capacitor bank discharge through a solenoid field coil (located either inside the gun inner electrode or outside the outer gun electrode) and the associated diffusion of magnetic flux through neighboring conductors, (2) gas flow into the gun annular region from a simulated puffed gas valve plenum, (3) formation and motion of a current sheet produced by J x B forces resulting from discharge of the gun capacitor bank through the plasma load between the coaxial gun electrodes, (4) the subsequent stretching and reconnection of the poloidal field lines to form a compact torus plasma ring, and (5) finally the discharge of the accelerator capacitor bank producing an additional toroidal field for acceleration of the plasma ring. The code has been extended to include various models for gas breakdown, plasma anomalous resistivity, and mass entrainment from ablation of electrode material

  3. Experience curve for natural gas production by hydraulic fracturing

    International Nuclear Information System (INIS)

    Fukui, Rokuhei; Greenfield, Carl; Pogue, Katie; Zwaan, Bob van der

    2017-01-01

    From 2007 to 2012 shale gas production in the US expanded at an astounding average growth rate of over 50%/yr, and thereby increased nearly tenfold over this short time period alone. Hydraulic fracturing technology, or “fracking”, as well as new directional drilling techniques, played key roles in this shale gas revolution, by allowing for extraction of natural gas from previously unviable shale resources. Although hydraulic fracturing technology had been around for decades, it only recently became commercially attractive for large-scale implementation. As the production of shale gas rapidly increased in the US over the past decade, the wellhead price of natural gas dropped substantially. In this paper we express the relationship between wellhead price and cumulative natural gas output in terms of an experience curve, and obtain a learning rate of 13% for the industry using hydraulic fracturing technology. This learning rate represents a measure for the know-how and skills accumulated thus far by the US shale gas industry. The use of experience curves for renewable energy options such as solar and wind power has allowed analysts, practitioners, and policy makers to assess potential price reductions, and underlying cost decreases, for these technologies in the future. The reasons for price reductions of hydraulic fracturing are fundamentally different from those behind renewable energy technologies – hence they cannot be directly compared – and hydraulic fracturing may soon reach, or maybe has already attained, a lower bound for further price reductions, for instance as a result of its water requirements or environmental footprint. Yet, understanding learning-by-doing phenomena as expressed by an industry-wide experience curve for shale gas production can be useful for strategic planning in the gas sector, as well as assist environmental policy design, and serve more broadly as input for projections of energy system developments. - Highlights: • Hydraulic

  4. Disruption mitigation experiment with massive gas injection of HT-7

    International Nuclear Information System (INIS)

    Zhuang Huidong; Zhang Xiaodong

    2013-01-01

    Massive gas injection (MGI) is a promising method on disruption mitigation. The working principle of the fast valve for disruption mitigation was introduced. The disruption mitigation experiments by MGI on HT-7 were described. The experiment shows that the impurities radiation is improved by injecting appropriate amount of gas, and the current quench rate is slow down, so the electromagnetic load on the device is mitigated. The experiments show that the fast valve can completely satisfy the requirement of disruption mitigation on HT-7. (authors)

  5. Modeling of modification experiments involving neutral-gas release

    International Nuclear Information System (INIS)

    Bernhardt, P.A.

    1983-01-01

    Many experiments involve the injection of neutral gases into the upper atmosphere. Examples are critical velocity experiments, MHD wave generation, ionospheric hole production, plasma striation formation, and ion tracing. Many of these experiments are discussed in other sessions of the Active Experiments Conference. This paper limits its discussion to: (1) the modeling of the neutral gas dynamics after injection, (2) subsequent formation of ionosphere holes, and (3) use of such holes as experimental tools

  6. Experimental evaluation of blockage ratio and plenum evacuation system flow effects on pressure distribution for bodies of revolution in 0.1 scale model test section of NASA Lewis Research Center's proposed altitude wind tunnel

    Science.gov (United States)

    Burley, Richard R.; Harrington, Douglas E.

    1987-01-01

    An experimental investigation was conducted in the slotted test section of the 0.1-scale model of the proposed Altitude Wind Tunnel to evaluate wall interference effects at tunnel Mach numbers from 0.70 to 0.95 on bodies of revolution with blockage rates of 0.43, 3, 6, and 12 percent. The amount of flow that had to be removed from the plenum chamber (which surrounded the slotted test section) by the plenum evacuation system (PES) to eliminate wall interference effects was determined. The effectiveness of tunnel reentry flaps in removing flow from the plenum chamber was examined. The 0.43-percent blockage model was the only one free of wall interference effects with no PES flow. Surface pressures on the forward part of the other models were greater than interference-free results and were not influenced by PES flow. Interference-free results were achieved on the aft part of the 3- and 6-percent blockage models with the proper amount of PES flow. The required PES flow was substantially reduced by opening the reentry flaps.

  7. Effects of particle exhaust on neutral compression ratios in DIII-D

    International Nuclear Information System (INIS)

    Colchin, R.J.; Maingi, R.; Wade, M.R.; Allen, S.L.; Greenfield, C.M.

    1998-08-01

    In this paper, neutral particles in DIII-D are studied via their compression in the plenum and via particle exhaust. The compression of gas in the plena is examined in terms of the magnetic field configuration and wall conditions. DIII-D compression ratios are observed in the range from 1 to ≥ 1,000. Particle control ultimately depends on the exhaust of neutrals via plenum or wall pumping. Wall pumping or outgassing is calculated by means of a detailed particle balance throughout individual discharges, and its effect on particle control is discussed. It is demonstrated that particle control through wall conditioning leads to lower normalized densities. A two-region model shows that the gas compression ratio (C div = divertor plenum neutral pressure/torus neutral pressure) can be interpreted in relation to gas flows in the torus and divertor including the pumping speed of the plenum cryopumps, plasma pumping, and the pumping or outgassing of the walls

  8. Axial transport of fission gas in LWR fuel rods

    International Nuclear Information System (INIS)

    Kinoshita, M.

    1983-01-01

    With regard to fission gas transportation inside the fuel rod, the following three mechanisms are important: (1) a localized and time dependent fission gas release from UO 2 fuel to pellet/clad gap, (2) the consequent gas pressure difference between the gap and the plenum, and (3) the inter-diffusion of initially filled Helium and released fission gas such as Xenon. Among these three mechanisms, the 2nd mechanism would result in the one dimensional flow through P/C gap in the axial direction, while the 3rd would average the local fission gas concentration difference. In this paper, an attempt was made to develop a computerized model, LINUS (LINear flow and diffusion under Un-Steady condition) describing the above two mechanisms, items (2) and (3). The item (1) is treated as an input. The code was applied to analyse short length experimental fuel rods and long length commercial fuel rods. The calculated time evolution of Xe concentration along the fuel column shows that the dilution rate of Xe in commercial fuel rods is much slower than that in short experimental fuel rods. Some other sensitivity studies, such as the effect of pre-pressurization, are also presented. (author)

  9. Study on cooling model for debris in lower plenum and countermeasures for prevention of focusing effect

    International Nuclear Information System (INIS)

    Guan Zhonghua; Yu Hongxing; Jiang Guangming

    2008-01-01

    From the basic energy conservation equations and experimental or empirical correlations, an intact model is constructed for the thermal calculation of the core debris in the lower plenum. For verification of this model, the results of two calculations for AP600 and AP1000 plants are compared with those presented in relevant literature. The analysis highlights on the impact of the decay heat power density and the focusing effect. In order to mitigate the focusing effect, it is proposed in this paper to change the lower head profile from hemisphere to parabola. The results show that this change of lower head profile can change the heat flux distribution of the debris, and mitigate the focusing effect. (authors)

  10. The characterization and monitoring of metallic fuel breaches in EBR-2

    International Nuclear Information System (INIS)

    Pahl, R.G.; Batte, G.L.; Mikaili, R.; Lambert, J.D.B.; Hofman, G.L.

    1991-01-01

    This paper discusses the characterization and monitoring of metallic fuel breaches which is now a significant part of the Integral Fast Reactor fuel testing program at Argonne National Laboratory. Irradiation experience with failed metallic fuel now includes natural breaches in the plenum and fuel column regions in lead ''endurance'' tests as well as fuel column breaches in artificially-defected fuel which have operated for months in the run-beyond-cladding breach (RBCB) mode. Analyses of the fission gas (FG) release-to-birth (R/B) ratios of selected historical breaches have been completed and have proven to be very useful in differentiating between plenum and fuel column breaches

  11. Ben Macdhui High Altitude Trace Gas and Aerosol Transport Experiment

    CSIR Research Space (South Africa)

    Piketh, SJ

    1999-01-01

    Full Text Available The Ben Macdhui High Altitude Aerosol and Trace Gas Transport Experiment (BHATTEX) was started to characterize the nature and magnitude of atmospheric, aerosol and trace gas transport paths recirculation over and exiting from southern Africa...

  12. Tag gas burnup based on three-dimensional FTR analysis

    International Nuclear Information System (INIS)

    Kidman, R.B.

    1976-01-01

    Flux spectra from a three-dimensional diffusion theory analysis of the Fast Test Reactor (FTR) are used to predict gas tag ratio changes, as a function of exposure, for each FTR fuel and absorber subassembly plenum. These flux spectra are also used to predict Xe-125 equilibrium activities in absorber plena in order to assess the feasibility of using Xe-125 gamma rays to detect and distinguish control rod failures from fuel rod failures. Worst case tag burnup changes are used in conjunction with burnup and mass spectrometer uncertainties to establish the minimum spacing of tags which allows the tags to be unambiguously identified

  13. Analysis of SONACO axial cooling experiments

    International Nuclear Information System (INIS)

    Sigg, B.; Dury, T.V.; Hudina, M.

    1994-01-01

    The SONACO test rig contained a sodium-cooled, electrically heated 37-pin bundle. On this rig, a series of forced, mixed and natural convection experiments have been performed with the aim of contributing to the understanding of thermal-hydraulic phenomena and providing data for code validation for a subassembly at decay heat power level with low flow or stagnant coolant. The test section and especially the heater pins were equipped with an extensive number of chromel-alumel thermocouples. In addition, special permanent-magnet probes were used for measuring local velocities. In this paper we give a survey of results from axial cooling experiments, where heat was removed by natural convection to a cooling coil situated in the coolant channel (plenum) above the bundle. The experimental conditions led to turbulent convection with a slowly varying, large scale flow pattern. It is shown that a power tilt in the bundle reduces these fluctuations but does not eliminate them. For the uniformly heated bundle, aglebraic expressions for the average turbulent heat flux as well as for temperature and velocity fluctuations are derived from a second-moments model and compared with experimental data. Furthermore, heat transfer in the plenum and the consequences of the SONACO experiments for the coolability of reactor fuel elements under loss-of-flow conditions are discussed. ((orig.))

  14. Components inspection of Monju, a sodium bonded type control rod

    International Nuclear Information System (INIS)

    Harada, Kiyoshi; Matsushita, Yuichi; Lee, Chunchan; Abe, Hideaki; Watahiki, Naohisa

    2002-03-01

    This Report addresses a result of a sodium test conducted on components of a Double Poral Filter Sodium Bonded Type Control Rod that is expected to be a next generation, long life Control Rod. Upper and lower Poral Filter Sodium Bonded Type Control Rod components were mocked up to conduct a sodium test. During the test, sodium chargeability, formation of Gas Plenum at the upper part of the components, sodium drain-ability and NaOH clean-ability were recognized under actual plant condition. The following are results obtained: (1) Sodium Chargeability at Control Rod Insertion to EVST. Sodium was charged into the components when the mocked-up was inserted in sodium of 190degC, with insertion speed of 6 m/min which is an actual insertion speed to EVST. (2) Formation of Upper Gas Plenum by Helium Gas generated in Control Rod Components Gas Plenum formation within deviation of 9% was confirmed by releasing helium gas into the mocked-up which is immersed in sodium of 620degC and 190degC. Length of Gas Plenum is confirmed to be retained in certain length even if helium gas is further released into formed Gas Plenum. (3) Sodium Drain-ability of Control Rod Components when Drawing from EVST. Drain-ability was confirmed to be sufficient and no sodium residue was found in the mocked-up when the mocked-up was drawn out from sodium of 190degC, with drawing speed of 6 m/min which is an actual drawing speed from EVST. (4) Clean-ability of NaOH Solution against Sodium Residue in Control Rod Components. Sodium and NaOH solution reacted calmly, however, clean-ability was not sufficient. When Sodium fully remained in Control Rod Components, it made circulation of NaOH solution not enough. (author)

  15. Mechanical energy release in CABRI-2 experiments with Viggen-4 fuel pins

    International Nuclear Information System (INIS)

    Wolff, J.

    1993-07-01

    The results of mechanical energy release evaluations in CABRI-2 experiments with Viggen-4 fuel pins (12 atom % burnup) are described. In general the experience gained by the CABRI-1 experiments is confirmed. Those physical phenomena are enhanced which are influenced by the release of fission products. Especially the late blow-out of pressurized fission gases from the lower test pin plenum led to large flow variations. The corresponding mechanical power releases are low

  16. Natural gas market assessment. Canadian natural gas market mechanisms: Recent experiences and developments

    International Nuclear Information System (INIS)

    1993-11-01

    The increase in natural gas demand and the associated expansions of most of the pipeline systems serving western Canada have reduced the excess deliverability or excess productive capacity that existed at the time of deregulation of the natural gas industry in 1985. Based on an industry survey, the responses of natural gas buyers and sellers to recent supply difficulties are described. Specific production, transportation, and contractual difficulties were encountered in winter 1992/93 as production was stretched to meet record levels of demand during periods of very cold temperatures and as short-term spot prices reached very high levels. Problems at this time included wellhead freezeups, pipeline outages, and inadequate contract terms and conditions. Methods used to maintain gas flows to end users are reviewed, including a discussion of force majeure, spot gas purchases, storage, supply curtailment, and special loan arrangements. In 1992/93, in most instances where the responsibility fell on the end-user to solve the supply problem, the difficulty was shifted to local distribution companies who have traditionally had more experience with such situations. No cases were identified where either a firm or interruptible end-user was forced to curtail gas consumption because of inadequate supply. New market mechanisms are emerging that will enable buyers and sellers of western Canadian gas to avoid many of the problems encountered in 1992/93. These include prearranged backstopping arrangements, short-term spot markets, access to other gas basins, standardized gas contracts, electronic trading, and price risk management tools. 11 figs

  17. A system for the discharge of gas bubbles from the coolant flow of a nuclear reactor cooled by forced circulation

    International Nuclear Information System (INIS)

    Markfort, D.; Kaiser, A.; Dohmen, A.

    1975-01-01

    In a reactor cooled by forced circulation the gas bubbles carried along with the coolant flow are separated before entering the reactor core or forced away into the external zones. For this purpose the coolant is radially guided into a plenum below the core and deflected to a tangential direction by means of flow guide elements. The flow runs spirally downwards. On the bubbles, during their dwell time in this channel, the buoyant force and a force towards the axis of symmetry of the tank are exerted. The major part of the coolant is directed into a radial direction by means of a guiding apparatus in the lower section of the channel and guided through a chimney in the plenum to the center of the reactor core. This inner chimney is enclosed by an outer chimney for the core edge zones through which coolant with a small share of bubbles is taken away. (RW) [de

  18. Current collector design for closed-plenum polymer electrolyte membrane fuel cells

    Science.gov (United States)

    Daniels, F. A.; Attingre, C.; Kucernak, A. R.; Brett, D. J. L.

    2014-03-01

    This work presents a non-isothermal, single-phase, three-dimensional model of the effects of current collector geometry in a 5 cm2 closed-plenum polymer electrolyte membrane (PEM) fuel cell constructed using printed circuit boards (PCBs). Two geometries were considered in this study: parallel slot and circular hole designs. A computational fluid dynamics (CFD) package was used to account for species, momentum, charge and membrane water distribution within the cell for each design. The model shows that the cell can reach high current densities in the range of 0.8 A cm-2-1.2 A cm-2 at 0.45 V for both designs. The results indicate that the transport phenomena are significantly governed by the flow field plate design. A sensitivity analysis on the channel opening ratio shows that the parallel slot design with a 50% opening ratio shows the most promising performance due to better species, heat and charge distribution. Modelling and experimental analysis confirm that flooding inhibits performance, but the risk can be minimised by reducing the relative humidity of the cathode feed to 50%. Moreover, overheating is a potential problem due to the insulating effect of the PCB base layer and as such strategies should be implemented to combat its adverse effects.

  19. Modeling gas migration experiments in repository host rocks for the MEGAS project

    International Nuclear Information System (INIS)

    Worgan, K.; Impey, M.; Volckaert, G.; DePreter, P.

    1993-01-01

    In response to concerns over the possibility of hydrogen gas generation within an underground repository for high-level radioactive waste, and its implications for repository safety, a joint European research study (MEGAS) is underway. Its aims are to understand and characterize the behavior of gas migration within an argillacious, host-rock. Laboratory experiments are being carried out by SCK/CEN, BGS and ISMES. SCK/CEN are also conducting in situ experiments at the underground laboratory at Mol, Belgium. Modeling of gas migration is being done in parallel with the experiments, by Intera Information Technologies. A two-phase flow code, TOPAZ, has been developed specifically for this work. In this paper the authors report on the results of some preliminary calculations performed with TOPAZ, in advance of the in situ experiments

  20. Coupled CFD - system-code simulation of a gas cooled reactor

    International Nuclear Information System (INIS)

    Yan, Yizhou; Rizwan-uddin

    2011-01-01

    A generic coupled CFD - system-code thermal hydraulic simulation approach was developed based on FLUENT and RELAP-3D, and applied to LWRs. The flexibility of the coupling methodology enables its application to advanced nuclear energy systems. Gas Turbine - Modular Helium Reactor (GT-MHR) is a Gen IV reactor design which can benefit from this innovative coupled simulation approach. Mixing in the lower plenum of the GT-MHR is investigated here using the CFD - system-code coupled simulation tool. Results of coupled simulations are presented and discussed. The potential of the coupled CFD - system-code approach for next generation of nuclear power plants is demonstrated. (author)

  1. Combustor nozzles in gas turbine engines

    Science.gov (United States)

    Johnson, Thomas Edward; Keener, Christopher Paul; Stewart, Jason Thurman; Ostebee, Heath Michael

    2017-09-12

    A micro-mixer nozzle for use in a combustor of a combustion turbine engine, the micro-mixer nozzle including: a fuel plenum defined by a shroud wall connecting a periphery of a forward tube sheet to a periphery of an aft tubesheet; a plurality of mixing tubes extending across the fuel plenum for mixing a supply of compressed air and fuel, each of the mixing tubes forming a passageway between an inlet formed through the forward tubesheet and an outlet formed through the aft tubesheet; and a wall mixing tube formed in the shroud wall.

  2. Injector design for liner-on-target gas-puff experiments

    Science.gov (United States)

    Valenzuela, J. C.; Krasheninnikov, I.; Conti, F.; Wessel, F.; Fadeev, V.; Narkis, J.; Ross, M. P.; Rahman, H. U.; Ruskov, E.; Beg, F. N.

    2017-11-01

    We present the design of a gas-puff injector for liner-on-target experiments. The injector is composed of an annular high atomic number (e.g., Ar and Kr) gas and an on-axis plasma gun that delivers an ionized deuterium target. The annular supersonic nozzle injector has been studied using Computational Fluid Dynamics (CFD) simulations to produce a highly collimated (M > 5), ˜1 cm radius gas profile that satisfies the theoretical requirement for best performance on ˜1-MA current generators. The CFD simulations allowed us to study output density profiles as a function of the nozzle shape, gas pressure, and gas composition. We have performed line-integrated density measurements using a continuous wave (CW) He-Ne laser to characterize the liner gas density. The measurements agree well with the CFD values. We have used a simple snowplow model to study the plasma sheath acceleration in a coaxial plasma gun to help us properly design the target injector.

  3. Radial midframe baffle for can-annular combustor arrangement having tangentially oriented combustor cans

    Science.gov (United States)

    Rodriguez, Jose L.

    2015-09-15

    A can-annular gas turbine engine combustion arrangement (10), including: a combustor can (12) comprising a combustor inlet (38) and a combustor outlet circumferentially and axially offset from the combustor inlet; an outer casing (24) defining a plenum (22) in which the combustor can is disposed; and baffles (70) configured to divide the plenum into radial sectors (72) and configured to inhibit circumferential motion of compressed air (16) within the plenum.

  4. High pressure gas spheres for neutron and photon experiments

    Science.gov (United States)

    Rupp, G.; Petrich, D.; Käppeler, F.; Kaltenbaek, J.; Leugers, B.; Reifarth, R.

    2009-09-01

    High pressure gas spheres have been designed and successfully used in several nuclear physics experiments on noble gases. The pros and cons of this solution are the simple design and the high reliability versus the fact that the density is limited to 40-60% of liquid or solid gas samples. Originally produced for neutron capture studies at keV energies, the comparably small mass of the gas spheres were an important advantage, which turned out to be of relevance for other applications as well. The construction, performance, and operation of the spheres are described and examples for their use are presented.

  5. Reflooding phenomena of German PWR estimated from CCTF [Cylindrical Core Test Facility], SCTF [Slab Core Test Facility] and UPTF [Upper Plenum Test Facility] results

    International Nuclear Information System (INIS)

    Murao, Y.; Iguchi, T.; Sugimoto, J.

    1988-09-01

    The reflooding behavior in a PWR with a combined injection type ECCS was studied by comparing the test results from Cylindrical Core Test Facility (CCTF), Slab Core Test Facility (SCTF) and Upper Plenum Test Facility (UPTF). Core thermal-hydraulics is discussed mainly based on SCTF test data. In addition, the water accumulation behavior in hot legs and the break-through characteristics at tie plate are discussed

  6. Sounding experiments of high pressure gas discharge

    International Nuclear Information System (INIS)

    Biele, Joachim K.

    1998-01-01

    A high pressure discharge experiment (200 MPa, 5·10 21 molecules/cm 3 , 3000 K) has been set up to study electrically induced shock waves. The apparatus consists of the combustion chamber (4.2 cm 3 ) to produce high pressure gas by burning solid propellant grains to fill the electrical pump chamber (2.5 cm 3 ) containing an insulated coaxial electrode. Electrical pump energy up to 7.8 kJ at 10 kV, which is roughly three times of the gas energy in the pump chamber, was delivered by a capacitor bank. From the current-voltage relationship the discharge develops at rapidly decreasing voltage. Pressure at the combustion chamber indicating significant underpressure as well as overpressure peaks is followed by an increase of static pressure level. These data are not yet completely understood. However, Lorentz forces are believed to generate pinching with subsequent pinch heating, resulting in fast pressure variations to be propagated as rarefaction and shock waves, respectively. Utilizing pure axisymmetric electrode initiation rather than often used exploding wire technology in the pump chamber, repeatable experiments were achieved

  7. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 1. Numerical investigation for the rationalization of hydrodynamics in the upper plenum

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Yamaguchi, Akira

    2002-02-01

    A large-scale sodium-cooled fast breeder reactor in feasibility studies on commercialized fast reactors has a tendency of consideration of thorough simplified and compacted system designs to realize drastic economical improvements. Therefore, special attention should be paid to thermohydraulic designs for a gas entrainment behavior from free surfaces, a flow-induced vibration of in-vessel components, a thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermohydraulic characteristics in the upper plenum, and to investigate suitable in-vessel structure for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) It is difficult to rationale in-vessel flow patterns through adjustments of porous ratio and pressure loss for a hold down plate and baffle plates installed in an upper core structure. (2) Dummy plug insertion to a slit of the upper core structure is one of effective measures to stabilize in-vessel flow patterns. (3) Flow guide devices such as a baffle ring and a partial inner barrel are also effective measures to eliminate impinging jet to a dipped plate (D/P) and to reduce horizontal flow velocity components at free surface. (4) Installations of labyrinth structures to a R/V - D/P gap is successful for decreasing of free surface horizontal flows. (5) Gap closing of an in-vessel fuel pot and two cold trap components has the effects of reductions for free surface horizontal flows and for the difference of free surface levels. Following future investigations are important preventive measures against the gas entrainment from the free surface. (1) Flattening of spatial axial velocity distributions at the R/V - D/P gap. (2) Alleviation measures of vortex concentration at free surface. (3) Separation measures of 3-dimensional vortex

  8. Investigation of physico-chemical processes in hypervelocity MHD-gas acceleration wind tunnels

    International Nuclear Information System (INIS)

    Alfyorov, V.I.; Dmitriev, L.M.; Yegorov, B.V.; Markachev, Yu.E.

    1995-01-01

    The calculation results for nonequilibrium physicochemical processes in the circuit of the hypersonic MHD-gas acceleration wind tunnel are presented. The flow in the primary nozzle is shown to be in thermodynamic equilibrium at To=3400 K, Po=(2∼3)x10 5 Pa, M=2 used in the plenum chamber. Variations in the static pressure due to oxidation reaction of Na, K are pointed out. The channels of energy transfer from the electric field to different degrees of freedom of an accelerated gas with Na, K seeds are considered. The calculation procedure for gas dynamic and kinetic processes in the MHD-channel using measured parameters is suggested. The calculated results are compared with the data obtained in a thermodynamic gas equilibrium assumption. The flow in the secondary nozzle is calculated under the same assumptions and the gas parameters at its exit are evaluated. Particular attention is given to the influence of seeds on flows over bodies. It is shown that the seeds exert a very small influence on the flow behind a normal shock wave. The seeds behind an oblique shock wave accelerate deactivation of vibrations of N 2 , but this effect is insignificant

  9. Partner Country Series: Gas Pricing - China's Challenges and IEA Experience

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-07-01

    China will play a positive role in the global development of gas, the International Energy Agency’s (IEA) Executive Director, Maria Van der Hoeven has said in Beijing on 11 September, 2012 when launching a new IEA report: Gas Pricing and Regulation, China’s challenges and IEA experiences. In line with its aim to meet growing energy demand while shifting away from coal, China has set an ambitious goal of doubling its use of natural gas from 2011 levels by 2015. Prospects are good for significant new supplies – both domestic and imported, conventional and unconventional – to come online in the medium-term, but notable challenges remain, particularly concerning gas pricing and the institutional and regulatory landscape. While China’s circumstances are, in many respects unique, some current issues are similar to those a number of IEA countries have faced. This report highlights some key challenges China faces in its transition to greater reliance on natural gas, then explores in detail relevant experiences from IEA countries, particularly in the United Kingdom, the Netherlands, and the United States as well as the European Union (EU). Preliminary suggestions about how lessons learned in other countries could be applied to China’s situation are offered as well. The aim of this report is to provide stakeholders in China with a useful reference as they consider decisions about the evolution of the gas sector in their country.

  10. Computational Fluid Dynamic Analysis of the VHTR Lower Plenum Standard Problem

    International Nuclear Information System (INIS)

    Johnson, Richard W.; Schultz, Richard R.

    2009-01-01

    The United States Department of Energy is promoting the resurgence of nuclear power in the U. S. for both electrical power generation and production of process heat required for industrial processes such as the manufacture of hydrogen for use as a fuel in automobiles. The DOE project is called the next generation nuclear plant (NGNP) and is based on a Generation IV reactor concept called the very high temperature reactor (VHTR), which will use helium as the coolant at temperatures ranging from 450 C to perhaps 1000 C. While computational fluid dynamics (CFD) has not been used for past safety analysis for nuclear reactors in the U.S., it is being considered for safety analysis for existing and future reactors. It is fully recognized that CFD simulation codes will have to be validated for flow physics reasonably close to actual fluid dynamic conditions expected in normal and accident operational situations. To this end, experimental data have been obtained in a scaled model of a narrow slice of the lower plenum of a prismatic VHTR. The present report presents results of CFD examinations of these data to explore potential issues with the geometry, the initial conditions, the flow dynamics and the data needed to fully specify the inlet and boundary conditions; results for several turbulence models are examined. Issues are addressed and recommendations about the data are made

  11. Experience transfer in Norwegian oil and gas industry: Approaches and organizational mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Aase, Karina

    1997-12-31

    The main objective of this thesis has been to explore how experience transfer works in Norwegian oil and gas industry. This includes how the concept of experience transfer is defined, what the barriers to achieve experience transfer are, how the oil and gas companies address experience transfer, and how these approaches work. The thesis is organized in five papers: (1) describes how organizational members perceive experience transfer and then specifies the organizational and structural barriers that must be overcome to achieve efficient transfer. (2) discusses the organizational means an oil company implements to address experience transfer. (3) describes a process of improving and using requirement and procedure handbooks for experience transfer. (4) explores how the use of information technology influences experience transfer. (5) compares organizational members` perceptions of experience transfer means in an oil company and an engineering company involved in offshore development projects. 277 refs., 3 figs., 29 tabs.

  12. Experience transfer in Norwegian oil and gas industry: Approaches and organizational mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Aase, Karina

    1998-12-31

    The main objective of this thesis has been to explore how experience transfer works in Norwegian oil and gas industry. This includes how the concept of experience transfer is defined, what the barriers to achieve experience transfer are, how the oil and gas companies address experience transfer, and how these approaches work. The thesis is organized in five papers: (1) describes how organizational members perceive experience transfer and then specifies the organizational and structural barriers that must be overcome to achieve efficient transfer. (2) discusses the organizational means an oil company implements to address experience transfer. (3) describes a process of improving and using requirement and procedure handbooks for experience transfer. (4) explores how the use of information technology influences experience transfer. (5) compares organizational members` perceptions of experience transfer means in an oil company and an engineering company involved in offshore development projects. 277 refs., 3 figs., 29 tabs.

  13. In situ water and gas injection experiments performed in the Hades Underground Research Facility

    Energy Technology Data Exchange (ETDEWEB)

    Volckaert, G.; Ortiz, L.; Put, M. [SCK-CEN, Mol (Belgium). Geological Waste Disposal Unit

    1995-12-31

    The movement of water and gas through plastic clay is an important subject in the research at SCK-CEN on the possible disposal of high level radioactive waste in the Boom clay layer at Mol. Since the construction of the Hades underground research facility in 1983, SCK-CEN has developed and installed numerous piezometers for the geohydrologic characterization and for in situ radionuclide migration experiments. In situ gas and water injection experiments have been performed at two different locations in the underground laboratory. The first location is a multi filter piezometer installed vertically at the bottom of the shaft in 1986. The second location is a three dimensional configuration of four horizontal multi piezometers installed from the gallery. This piezometer configuration was designed for the MEGAS (Modelling and Experiments on GAS migration through argillaceous rocks) project and installed in 1992. It contains 29 filters at distances between 10 m and 15 m from the gallery in the clay. Gas injection experiments show that gas breakthrough occurs at a gas overpressure of about 0.6 MPa. The breakthrough occurs by the creation of gas pathways along the direction of lowest resistance i.e. the zone of low effective stress resulting from the drilling of the borehole. The water injections performed in a filter -- not used for gas injection -- show that the flow of water is also influenced by the mechanical stress conditions. Low effective stress leads to higher hydraulic conductivity. However, water overpressures up to 1.3 MPa did not cause hydrofracturing. Water injections performed in a filter previously used for gas injections, show that the occluded gas hinders the water flow and reduces the hydraulic conductivity by a factor two.

  14. French gas cooled reactor experience with moisture ingress

    International Nuclear Information System (INIS)

    Bastien, D.; Brie, M.

    1995-01-01

    During the history of operation of six gas cooled reactors in France, some experience has been gained with accidental water ingress into the primary system. This occurred as a result of leaks in steam generators. This paper describes the cause of the leaks, and the resulting consequences. (author). 2 refs, 8 figs

  15. Stimulating shale gas development in China: A comparison with the US experience

    International Nuclear Information System (INIS)

    Tian, Lei; Wang, Zhongmin; Krupnick, Alan; Liu, Xiaoli

    2014-01-01

    In this paper, we use the US shale gas experience to shed light on how China might overcome the innovation problem inherent in exploring and developing shale gas plays with complex geology. We separate shale gas development into two stages, an innovation stage and a scaling-up stage, with the first presenting a much bigger challenge than the latter. Our analysis suggests that China's national oil companies offer the best hope for overcoming the innovation problem. China's policy of opening shale gas development to new entrants is a market-oriented reform that can be justified on various grounds, but the new entrants will not play a major role in overcoming the innovation problem even though they may help scale up production later on. - Highlights: • We separate shale gas development into an innovation stage and a scaling up stage. • We use the US experience to elucidate how China may solve the innovation problem. • China has to rely on its national oil companies to solve the innovation problem. • Opening shale gas development to new entrants can be justified on several grounds. • New entrants will not help overcome the innovation problem

  16. A gas trapping method for high-throughput metabolic experiments.

    Science.gov (United States)

    Krycer, James R; Diskin, Ciana; Nelson, Marin E; Zeng, Xiao-Yi; Fazakerley, Daniel J; James, David E

    2018-01-01

    Research into cellular metabolism has become more high-throughput, with typical cell-culture experiments being performed in multiwell plates (microplates). This format presents a challenge when trying to collect gaseous products, such as carbon dioxide (CO2), which requires a sealed environment and a vessel separate from the biological sample. To address this limitation, we developed a gas trapping protocol using perforated plastic lids in sealed cell-culture multiwell plates. We used this trap design to measure CO2 production from glucose and fatty acid metabolism, as well as hydrogen sulfide production from cysteine-treated cells. Our data clearly show that this gas trap can be applied to liquid and solid gas-collection media and can be used to study gaseous product generation by both adherent cells and cells in suspension. Since our gas traps can be adapted to multiwell plates of various sizes, they present a convenient, cost-effective solution that can accommodate the trend toward high-throughput measurements in metabolic research.

  17. Gas Detection for Experiments

    CERN Document Server

    Hay, D

    2001-01-01

    Flammable gases are often used in detectors for physics experiments. The storage, distribution and manipulation of such flammable gases present several safety hazards. As most flammable gases cannot be detected by human senses, specific well-placed gas detection systems must be installed. Following a request from the user group and in collaboration with CERN safety officers, risk analyses are performed. An external contractor, who needs to receive detailed user requirements from CERN, performs the installations. The contract is passed on a guaranteed results basis. Co-ordination between all the CERN groups and verification of the technical installation is done by ST/AA/AS. This paper describes and focuses on the structured methodology applied to implement such installations based on goal directed project management techniques (GDPM). This useful supervision tool suited to small to medium sized projects facilitates the task of co-ordinating numerous activities to achieve a completely functional system.

  18. Validation of mechanistic models for gas precipitation in solids during postirradiation annealing experiments

    Science.gov (United States)

    Rest, J.

    1989-12-01

    A number of different phenomenological models for gas precipitation in solids during postirradiation annealing experiments have been proposed. Validation of such mechanistic models for gas release and swelling is complicated by the use of data containing large systematic errors, and phenomena characterized by synergistic effects as well as uncertainties in materials properties. Statistical regression analysis is recommended for the selection of a reasonably well characterized data base for gas release from irradiated fuel under transient heating conditions. It is demonstrated that an appropriate data selection method is required in order to realistically examine the impact of differing descriptions of the phenomena, and uncertainties in selected materials properties, on the validation results. The results of the analysis show that the kinetics of gas precipitation in solids depend on bubble overpressurization effects and need to be accounted for during the heatup phase of isothermal heating experiments. It is shown that if only the total gas release values (as opposed to time-dependent data) were available, differentiation between different gas precipitation models would be ambiguous. The observed sustained increase in the fractional release curve at relatively high temperatures after the total precipitation of intragranular gas in fission gas bubbles is ascribed to the effects of a grain-growth/grain-boundary sweeping mechanism.

  19. Validation of mechanistic models for gas precipitation in solids during postirradiation annealing experiments

    International Nuclear Information System (INIS)

    Rest, J.

    1989-01-01

    A number of different phenomenological models for gas precipitation in solids during postirradiation annealing experiments have been proposed. Validation of such mechanistic models for gas release and swelling is complicated by the use of data containing large systematic errors, and phenomena characterized by synergistic effects as well as uncertainties in materials properties. Statistical regression analysis is recommended for the selection of a reasonably well characterized data base for gas release from irradiated fuel under transient heating conditions. It is demonstrated that an appropriate data selection method is required in order to realistically examine the impact of differing descriptions of the phenomena, and uncertainties in selected materials properties, on the validation results. The results of the analysis show that the kinetics of gas precipitation in solid depend on bubble overpressurization effects and need to be accounted for during the heatup phase of isothermal heating experiments. It is shown that if only the total gas release values (as opposed to time-dependent data) were available, differentiation between different gas precipitation models would be ambiguous. The observed sustained increase in the fractional release curve at relatively high temperatures after the total precipitation of intragranular gas in fission gas bubbles is ascribed to the effects of a grain-growth/grain-boundary sweeping mechanism. (orig.)

  20. Experience Transfer in Norwegian Oil and Gas Industry: Approaches and Organizational Mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Aase, Karina

    1997-07-01

    The core aim of the study is to explore the concept of experience transfer in oil and gas industry, and how an oil company approaches this concept. The thesis consists of five papers which are combined in a general description entitled 'Experience transfer in Norwegian oil and gas industry: approaches and organizational mechanisms'. The first paper describes how organizational members perceive experience transfer, and then specifies the many organizational and structural barriers that have to be overcome to achieve efficient experience transfer. The second paper elaborates and assesses the organizational means an oil company implements to address experience transfer. The third paper describes a process of improving and using requirement and procedure handbooks for experience transfer. The fourth paper explores in more detail how the use of information technology influences experience transfer. And the fifth paper compares organizational members' perceptions of experience transfer means in an oil company and an engineering company involved in offshore development projects. Some of the papers are based upon the same data material. Therefore there are reiterations in parts of the contents, especially in the methodological sections.

  1. Experience Transfer in Norwegian Oil and Gas Industry: Approaches and Organizational Mechanisms

    International Nuclear Information System (INIS)

    Aase, Karina

    1997-01-01

    The core aim of the study is to explore the concept of experience transfer in oil and gas industry, and how an oil company approaches this concept. The thesis consists of five papers which are combined in a general description entitled 'Experience transfer in Norwegian oil and gas industry: approaches and organizational mechanisms'. The first paper describes how organizational members perceive experience transfer, and then specifies the many organizational and structural barriers that have to be overcome to achieve efficient experience transfer. The second paper elaborates and assesses the organizational means an oil company implements to address experience transfer. The third paper describes a process of improving and using requirement and procedure handbooks for experience transfer. The fourth paper explores in more detail how the use of information technology influences experience transfer. And the fifth paper compares organizational members' perceptions of experience transfer means in an oil company and an engineering company involved in offshore development projects. Some of the papers are based upon the same data material. Therefore there are reiterations in parts of the contents, especially in the methodological sections

  2. Three Years of Experience of Wet Gas Allocation on Canyon Express

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Aditya; Hall, James; Letton, Winsor

    2005-07-01

    In September 2002, production was begun from the three fields that together form the Canyon Express System- King's Peak, Aconcagua, and Camden Hills. The 9 wells from these fields are connected to a pair of 12-inch flow lines carrying the commingled wet gas a distance of approximately 92 kilometers back to the Canyon Station platform for processing. At the 21st NSFMW in October 2003, an initial report was given on the status of Wet Gas Allocation for the Canyon Express project. As discussed in that paper, dual-differential, subsea wet gas meters were chosen for the task of allocating gas and liquids back to individual wells. However, since the gas from all three fields was very dry (Lockhart-Martinelli parameter less than 0.01) and because the operating pressures were quite high (250 bar), application of the dual-differential function of the meters yielded errors in both liquid and gas flow rates. Furthermore, as these problems were being uncovered, scale was beginning to collect inside some of the meters. Taken together, these problems produced system imbalances as great as 20%. To address the problems, one of the individual flow metering elements within each wet gas meter was chosen as the allocation meter, operating as a single-phase gas meter. After three years of operation of the Canyon Express Project, considerable experience has been accumulated. Since at the time it held the record for deep water hydrocarbon production, application of the technologies discussed here were challenging and required considerable flexibility. It is believed that the Canyon Express experiences will benefit future deep water flow metering projects. The knowledge acquired thus far is surveyed and summarized. The emphasis is on the technical aspects. (tk)

  3. Gas-filled hohlraum experiments at the national ignition facility.

    Energy Technology Data Exchange (ETDEWEB)

    Fernández, J. C. (Juan C.); Gautier, D. C. (Donald Cort); Goldman, S. R. (Sanford R.); Grimm, B. M.; Hegelich, B. M. (Bjorn M.); Kline, J. L. (John L.); Montgomery, D. S. (David S.); Lanier, N. E. (Nicholas E.); Rose, H. A. (Harvey A.); Schmidt, D. M. (David M.); Swift, D. C.; Workman, J. B. (Jonathan B.); Alvarez, Sharon; Bower, Dan.; Braun, Dave.; Campbell, K. (Katherine); DeWald, E.; Glenzer, S. (Siegfried); Holder, J. (Joe P.); Kamperschroer, J. H. (James H.); Kimbrough, Joe (Joseph R.); Kirkwood, Robert (Bob); Landen, O. L. (Otto L.); Mccarville, Tom (Tomas J.); Macgowan, B.; Mackinnon, A.; Niemann, C.; Schein, J.; Schneider, M; Watts, Phil; Young, Ben-li [number : znumber] 194154; Young B.

    2004-01-01

    The summary of this paper is: (1) We have fielded on NIF a gas-filled hohlraum designed for future ignition experiments; (2) Wall-motion measurements are consistent with LASNEX simulations; (3) LPI back-scattering results have confounded expectations - (a) Stimulated Brillouin (SBS) dominates Raman (SRS) for any gas-fill species, (b) Measured SBS time-averaged reflectivity values are high, peak values are even higher, (c) SRS and SBS peak while laser-pulse is rising; and (4) Plasma conditions at the onset of high back-scattering yield high SBS convective linear gain - Wavelengths of the back-scattered light is predicted by linear theory.

  4. Gas-filled hohlraum experiments at the national ignition facility

    International Nuclear Information System (INIS)

    Fernandez, J.C.; Gautier, D.C.; Goldman, S.R.; Grimm, B.M.; Hegelich, B.M.; Kline, J.L.; Montgomery, D.S.; Lanier, N.E.; Rose, H.A.; Schmidt, D.M.; Swift, D.C.; Workman, J.B.; Alvarez, Sharon; Bower, Dan; Braun, Dave; Campbell, K.; DeWald, E.; Glenzer, S.; Holder, J.; Kamperschroer, J.H.; Kimbrough, Joe; Kirkwood, Robert; Landen, O.L.; Mccarville, Tom; Macgowan, B.; Mackinnon, A.; Niemann, C.; Schein, J.; Schneider, M.; Watts, Phil; Young, Ben-li; Young B.

    2004-01-01

    The summary of this paper is: (1) We have fielded on NIF a gas-filled hohlraum designed for future ignition experiments; (2) Wall-motion measurements are consistent with LASNEX simulations; (3) LPI back-scattering results have confounded expectations - (a) Stimulated Brillouin (SBS) dominates Raman (SRS) for any gas-fill species, (b) Measured SBS time-averaged reflectivity values are high, peak values are even higher, (c) SRS and SBS peak while laser-pulse is rising; and (4) Plasma conditions at the onset of high back-scattering yield high SBS convective linear gain - Wavelengths of the back-scattered light is predicted by linear theory.

  5. Three-Dimensional Neutral Transport Simulations of Gas Puff Imaging Experiments

    International Nuclear Information System (INIS)

    Stotler, D.P.; DIppolito, D.A.; LeBlanc, B.; Maqueda, R.J.; Myra, J.R.; Sabbagh, S.A.; Zweben, S.J.

    2003-01-01

    Gas Puff Imaging (GPI) experiments are designed to isolate the structure of plasma turbulence in the plane perpendicular to the magnetic field. Three-dimensional aspects of this diagnostic technique as used on the National Spherical Torus eXperiment (NSTX) are examined via Monte Carlo neutral transport simulations. The radial width of the simulated GPI images are in rough agreement with observations. However, the simulated emission clouds are angled approximately 15 degrees with respect to the experimental images. The simulations indicate that the finite extent of the gas puff along the viewing direction does not significantly degrade the radial resolution of the diagnostic. These simulations also yield effective neutral density data that can be used in an approximate attempt to infer two-dimensional electron density and temperature profiles from the experimental images

  6. Rarefied gas electro jet (RGEJ) micro-thruster for space propulsion

    International Nuclear Information System (INIS)

    Blanco, Ariel; Roy, Subrata

    2017-01-01

    This article numerically investigates a micro-thruster for small satellites which utilizes plasma actuators to heat and accelerate the flow in a micro-channel with rarefied gas in the slip flow regime. The inlet plenum condition is considered at 1 Torr with flow discharging to near vacuum conditions (<0.05 Torr). The Knudsen numbers at the inlet and exit planes are ∼0.01 and ∼0.1, respectively. Although several studies have been performed in micro-hallow cathode discharges at constant pressure, to our knowledge, an integrated study of the glow discharge physics and resulting fluid flow of a plasma thruster under these low pressure and low Knudsen number conditions is yet to be reported. Numerical simulations of the charge distribution due to gas ionization processes and the resulting rarefied gas flow are performed using an in-house code. The mass flow rate, thrust, specific impulse, power consumption and the thrust effectiveness of the thruster are predicted based on these results. The ionized gas is modelled using local mean energy approximation. An electrically induced body force and a thermal heating source are calculated based on the space separated charge distribution and the ion Joule heating, respectively. The rarefied gas flow with these electric force and heating source is modelled using density-based compressible flow equations with slip flow boundary conditions. The results show that a significant improvement of specific impulse can be achieved over highly optimized cold gas thrusters using the same propellant. (paper)

  7. Autoresonant-spectrometric determination of the residual gas composition in the ALPHA experiment apparatus

    Energy Technology Data Exchange (ETDEWEB)

    Amole, C.; Capra, A.; Menary, S. [Department of Physics and Astronomy, York University, Toronto, Ontario M3J 1P3 (Canada); Ashkezari, M. D.; Hayden, M. E. [Department of Physics, Simon Fraser University, Burnaby, British Columbia V5A 1S6 (Canada); Baquero-Ruiz, M.; Chapman, S.; Little, A.; Povilus, A.; So, C.; Turner, M. [Department of Physics, University of California, Berkeley, California 94720-7300 (United States); Bertsche, W. [Department of Physics, Swansea University, Swansea SA2 8PP (United Kingdom); School of Physics and Astronomy, University of Manchester, Manchester M13 9PL, United Kingdom and The Cockcroft Institute, Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Butler, E. [Physics Department, CERN, CH-1211 Geneva 23 (Switzerland); Cesar, C. L.; Silveira, D. M. [Instituto de Fisica, Universidade Federal do Rio de Janeiro, Rio de Janeiro 21941-972 (Brazil); Charlton, M.; Eriksson, S.; Isaac, C. A.; Madsen, N.; Napoli, S. C. [Department of Physics, Swansea University, Swansea SA2 8PP (United Kingdom); Collaboration: ALPHA Collaboration; and others

    2013-06-15

    Knowledge of the residual gas composition in the ALPHA experiment apparatus is important in our studies of antihydrogen and nonneutral plasmas. A technique based on autoresonant ion extraction from an electrostatic potential well has been developed that enables the study of the vacuum in our trap. Computer simulations allow an interpretation of our measurements and provide the residual gas composition under operating conditions typical of those used in experiments to produce, trap, and study antihydrogen. The methods developed may also be applicable in a range of atomic and molecular trap experiments where Penning-Malmberg traps are used and where access is limited.

  8. Gas-grain simulation experiment module conceptual design and gas-grain simulation facility breadboard development

    Science.gov (United States)

    Zamel, James M.; Petach, Michael; Gat, Nahum; Kropp, Jack; Luong, Christina; Wolff, Michael

    1993-12-01

    This report delineates the Option portion of the Phase A Gas-Grain Simulation Facility study. The conceptual design of a Gas-Grain Simulation Experiment Module (GGSEM) for Space Shuttle Middeck is discussed. In addition, a laboratory breadboard was developed during this study to develop a key function for the GGSEM and the GGSF, specifically, a solid particle cloud generating device. The breadboard design and test results are discussed and recommendations for further studies are included. The GGSEM is intended to fly on board a low earth orbit (LEO), manned platform. It will be used to perform a subset of the experiments planned for the GGSF for Space Station Freedom, as it can partially accommodate a number of the science experiments. The outcome of the experiments performed will provide an increased understanding of the operational requirements for the GGSF. The GGSEM will also act as a platform to accomplish technology development and proof-of-principle experiments for GGSF hardware, and to verify concepts and designs of hardware for GGSF. The GGSEM will allow assembled subsystems to be tested to verify facility level operation. The technology development that can be accommodated by the GGSEM includes: GGSF sample generation techniques, GGSF on-line diagnostics techniques, sample collection techniques, performance of various types of sensors for environmental monitoring, and some off-line diagnostics. Advantages and disadvantages of several LEO platforms available for GGSEM applications are identified and discussed. Several of the anticipated GGSF experiments require the de-agglomeration and dispensing of dry solid particles into an experiment chamber. During the GGSF Phase A study, various techniques and devices available for the solid particle aerosol generator were reviewed. As a result of this review, solid particle de-agglomeration and dispensing were identified as key undeveloped technologies in the GGSF design. A laboratory breadboard version of a solid

  9. Current status of gas migration and swelling experiments using engineering scale model for immediate depth disposal in Japan

    International Nuclear Information System (INIS)

    Higashihara, Tomohiro; Ono, Makoto; Kawaragi, Chie; Saito, Shigeyuki

    2010-01-01

    In intermediate depth disposal facility of radioactive waste in Japan, waste is surrounded with bentonite layer to retard interaction of the waste and groundwater, because the bentonite layer saturated with the groundwater has very low hydraulic conductivity. On the other hand, it is important to confirm stability of barrier system for stress generated together with swelling of the bentonite and to understand effect of increase of gas pressure because of generation of hydrogen gas by corrosion of metallic waste. To understand and evaluate the swelling behavior of the bentonite layer, JNES carries out the experiment. In the experiments, we carry out the swelling experiment to examine the swelling behavior of the bentonite layer and the gas migration experiment to understand the gas migration behavior in the bentonite layer, using engineering scale model of the disposal facility. The swelling experiment has been in operation since June 2010. After this experiment, the gas migration experiment will start in July 2011. (orig.)

  10. Vented fuel experiment for gas-cooled fast reactor application

    International Nuclear Information System (INIS)

    Longest, A.W.; Gat, U.; Conlin, J.A.; Campana, R.J.

    1976-01-01

    A pressure-equalized and vented fuel rod is being irradiated in an instrumented capsule designated GB-10 to approximately 100MWd/kg-heavy metal. The fuel is a sol-gel-derived 88 at.% uranium (approximately 9% 235 U) and 12 at.% plutonium oxide, and the cladding is 20% cold-worked 316 stainless steel. The capsule is being irradiated in the Oak Ridge Research Reactor (ORR) and has exceeded a burnup of 70MWd/kg. The fuel has been operated at linear power rates of 39 and 44kW/m, and peak outer cladding temperature of 565 and 630 0 C respectively. A similar fuel rod in a previous capsule (GB-9) was subjected to 48kW/m (685 0 C). Helium gas sweeps through any portion of the three regions of the fuel rod, namely: fuel, blanket, and charcoal trap. The charcoal trap is operated at about 300 0 C. An on-line Ge(Li) detector is used to analyse release rates of several gamma-emitting noble gas isotopes. Analyses are performed primarily on sweep gas flowing through the entire fuel rod, and for sweeps over the top of the charcoal trap. Sweep gas samples are analyzed for stable noble gas isotopes. Results in the form of ratios of release rate over birth rate (R/B) and venting rate over birth rate (V/B) are derived. R/B rates range from 10 -4 % to 30% while V/B ranges from 10 -6 % to 30%. Flow conductance in the capsule was monitored by recording the flow rate and pressure drop across the fuel rod and inlet sweep line. The flow conductance has been falling with increasing burnup, currently restricting the flow to about 20ml (s.t.p.)/min at a pressure difference of about 1.5MPa. Venting rates of the gaseous fission products as a function of gas pressure in the range 6.9 to 1.4MPa have also been measured. Planned future experiments include the monitoring of tritium release, venting and cladding permeation rates, and its molecular form. First measurements have been made. A simulated leak experiment will determine the mixture of fission gases as a function of flow rate and the most

  11. SEASAT demonstration experiments with the offshore oil, gas and mining industries

    Science.gov (United States)

    Mourad, A. G.; Robinson, A. C.; Balon, J. E.

    1979-01-01

    Despite its failure, SEASAT-1 acquired a reasonable volume of data that can be used by industrial participants on a non-real-time basis to prove the concept of microwave sensing of the world's oceans from a satellite platform. The amended version of 8 experimental plans are presented, along with a description of the satellite, its instruments, and the data available. Case studies are summarized for the following experiments: (1) Beaufort Sea oil, gas, and Arctic operations; (2) Labrador Sea oil, gas, and sea ice; (3) Gulf of Mexico pipelines; (4) U.S. East Coast offshore oil and gas; (5) worldwide offshore drilling and production operations; (6) Equatorial East Pacific Ocean mining; (7) Bering Sea ice project; and (8) North Sea oil and gas.

  12. Daily Thermal Predictions of the AGR-1 Experiment with Gas Gaps Varying with Time

    Energy Technology Data Exchange (ETDEWEB)

    Grant Hawkes; James Sterbentz; John Maki; Binh Pham

    2012-06-01

    A new daily as-run thermal analysis was performed at the Idaho National Laboratory on the Advanced Gas Reactor (AGR) test experiment number one at the Advanced Test Reactor (ATR). This thermal analysis incorporates gas gaps changing with time during the irradiation experiment. The purpose of this analysis was to calculate the daily average temperatures of each compact to compare with experimental results. Post irradiation examination (PIE) measurements of the graphite holder and fuel compacts showed the gas gaps varying from the beginning of life. The control temperature gas gap and the fuel compact – graphite holder gas gaps were linearly changed from the original fabrication dimensions, to the end of irradiation measurements. A steady-state thermal analysis was performed for each daily calculation. These new thermal predictions more closely match the experimental data taken during the experiment than previous analyses. Results are presented comparing normalized compact average temperatures to normalized log(R/B) Kr-85m. The R/B term is the measured release rate divided by the predicted birth rate for the isotope Kr-85m. Correlations between these two normalized values are presented.

  13. Experience Transfer in Norwegian Oil and Gas Industry: Approaches and Organizational Mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Aase, Karina

    1997-07-01

    The core aim of the study is to explore the concept of experience transfer in oil and gas industry, and how an oil company approaches this concept. The thesis consists of five papers which are combined in a general description entitled 'Experience transfer in Norwegian oil and gas industry: approaches and organizational mechanisms'. The first paper describes how organizational members perceive experience transfer, and then specifies the many organizational and structural barriers that have to be overcome to achieve efficient experience transfer. The second paper elaborates and assesses the organizational means an oil company implements to address experience transfer. The third paper describes a process of improving and using requirement and procedure handbooks for experience transfer. The fourth paper explores in more detail how the use of information technology influences experience transfer. And the fifth paper compares organizational members' perceptions of experience transfer means in an oil company and an engineering company involved in offshore development projects. Some of the papers are based upon the same data material. Therefore there are reiterations in parts of the contents, especially in the methodological sections.

  14. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors

  15. Gas transport below artificial recharge ponds: insights from dissolved noble gases and a dual gas (SF6 and 3He) tracer experiment.

    Science.gov (United States)

    Clark, Jordan F; Hudson, G Bryant; Avisar, Dror

    2005-06-01

    A dual gas tracer experiment using sulfur hexafluoride (SF6) and an isotope of helium (3He) and measurements of dissolved noble gases was performed at the El Rio spreading grounds to examine gas transport and trapped air below an artificial recharge pond with a very high recharge rate (approximately 4 m day(-1)). Noble gas concentrations in the groundwater were greater than in surface water due to excess air formation showing that trapped air exists below the pond. Breakthrough curves of SF6 and 3He at two nearby production wells were very similar and suggest that nonequilibrium gas transfer was occurring between the percolating water and the trapped air. At one well screened between 50 and 90 m below ground, both tracers were detected after 5 days and reached a maximum at approximately 24 days. Despite the potential dilution caused by mixing within the production well, the maximum concentration was approximately 25% of the mean pond concentration. More than 50% of the SF6 recharged was recovered by the production wells during the 18 month long experiment. Our results demonstrate that at artificial recharge sites with high infiltration rates and moderately deep water tables, transport times between recharge locations and wells determined with gas tracer experiments are reliable.

  16. A gas puff experiment for partial simulation of compact toroid formation on MARAUDER

    International Nuclear Information System (INIS)

    Englert, S.E.; Englert, T.J.; Degnan, J.H.; Gahl, J.M.

    1994-01-01

    Preliminary results will be reported of a single valve gas puff experiment to determine spatial and spectral distribution of a gas during the early ionization stages. This experiment has been developed as a diagnostic test-bed for partial simulation of compact toroid formation on MARAUDER. The manner in which the experimental hardware has been designed allows for a wide range of diagnostic access to evaluate early time evolution of the ionization process. This evaluation will help contribute to a clearer understanding of the initial conditions for the formation stage of the compact toroid in the MARAUDER experiment, where 60 of the same puff valves are used. For the experiment, a small slice of the MARAUDER cylindrical gas injection and expansion region geometry have been re-created but in cartesian coordinates. All of the conditions in the experiment adhere as closely as possible to the MARAUDER experiment. The timing, current rise time, capacitance, resistance and inductance are appropriate to both the simulation of one of the 60 puff valves and current delivery to the load. Both time-resolved images and spectral data have been gathered for visible light emission of the plasma. Processed images reveal characteristics of spatial distribution of the current. Spectral data provide information with respect to electron temperature and density, and entrainment of contaminants

  17. Particle control in the DIII-D advanced divertor

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Lippmann, S.I.; Mahdavi, M.A.; Petrie, T.W.; Stambaugh, R.D.; Hogan, J.; Klepper, C.C.; Mioduszewski, P.; Owen, L.; Hill, D.N.; Rensink, M.; Buchenauer, D.

    1991-11-01

    A new, electrically biasable, semi-closed divertor was installed and operated in the D3-D lower outside divertor location. The semi-closed divertor has yielded static gas pressure buildups in the pumping plenum in excess of 10 mtorr. (The planned cryogenic pumping is not yet installed). Electrical bias controls the distribution of particle recycle between the inner and outer divertors by rvec E x rvec B drifts. Depending on sign, bias increases or decreases the plenum gas pressure. Bias greatly reduce the sensitivity of plenum pressure to separatrix position. In particular, rvec E x rvec B drifts in the D3-D geometry can direct plasma across a divertor target and then optimally into the pumping aperture. Bias, even without active pumping, has also demonstrated a limited control of ELMing H-mode plasma density. 5 refs., 8 figs

  18. Partner Country Series: Gas Pricing - China's Challenges and IEA Experience

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2012-07-01

    China will play a positive role in the global development of gas, the International Energy Agency’s (IEA) Executive Director, Maria Van der Hoeven has said in Beijing on 11 September, 2012 when launching a new IEA report: Gas Pricing and Regulation, China’s challenges and IEA experiences. In line with its aim to meet growing energy demand while shifting away from coal, China has set an ambitious goal of doubling its use of natural gas from 2011 levels by 2015. Prospects are good for significant new supplies – both domestic and imported, conventional and unconventional – to come online in the medium-term, but notable challenges remain, particularly concerning gas pricing and the institutional and regulatory landscape. While China’s circumstances are, in many respects unique, some current issues are similar to those a number of IEA countries have faced. This report highlights some key challenges China faces in its transition to greater reliance on natural gas, then explores in detail relevant experiences from IEA countries, particularly in the United Kingdom, the Netherlands, and the United States as well as the European Union (EU). Preliminary suggestions about how lessons learned in other countries could be applied to China’s situation are offered as well. The aim of this report is to provide stakeholders in China with a useful reference as they consider decisions about the evolution of the gas sector in their country.

  19. Gas-filled hohlraum experiments at the National Ignition Facility

    International Nuclear Information System (INIS)

    Fernandez, Juan C.; Goldman, S.R.; Kline, J.L.; Dodd, E.S.; Gautier, C.; Grim, G.P.; Hegelich, B.M.; Montgomery, D.S.; Lanier, N.E.; Rose, H.; Schmidt, D.W.; Workman, J.B.; Braun, D.G.; Dewald, E.L.; Landen, O.L.; Campbell, K.M.; Holder, J.P.; MacKinnon, A.J.; Niemann, C.; Schein, J.

    2006-01-01

    Experiments done at the National Ignition Facility laser [J. A. Paisner, E. M. Campbell, and W. Hogan, Fusion Technol. 26, 755 (1994)] using gas-filled hohlraums demonstrate a key ignition design feature, i.e., using plasma pressure from a gas fill to tamp the hohlraum-wall expansion for the duration of the laser pulse. Moreover, our understanding of hohlraum energetics and the ability to predict the hohlraum soft-x-ray drive has been validated in ignition-relevant conditions. Finally, the laser reflectivity from stimulated Raman scattering in the fill plasma, a key threat to hohlraum performance, is shown to be suppressed by choosing a design with a sufficiently high ratio of electron temperature to density

  20. Experiment on performance of upper head injection system with ROSA-II

    International Nuclear Information System (INIS)

    1978-05-01

    Of the total 10 ROSA-II/UHI performance tests, 6 were reported previously. The rest are presented and discussion is made on the effects of heat generation in the core and UHI injection and repeatability of experiments. In addition, the following are described: (1) Pressure spikes observed in the upper head after sudden stoppage of UHI injection, and (2) discharge flow oscillation possibly due to UHI water injection into the upper plenum. (auth.)

  1. Generation of the line radiation of argon added to DT gas in Iskra-5 experiments

    International Nuclear Information System (INIS)

    Bel'kov, S.A.; Bessarab, A.V.; Veselov, A.V.; Gaidash, V.A.; Dolgoleva, G.V.; Zhidkov, N.V.; Izgorodin, V.M.; Kirillov, G.A.; Kochemasov, G.G.; Litvin, D.N.; Martynenko, S.P.; Mitrofanov, E.I.; Murugov, V.M.; Mkhitar'yan, L.S.; Petrov, S.I.; Pinegin, A.V.; Punin, V.T.; Suslov, N.A.

    1998-01-01

    The first experiments measuring the density of a compressed deuterium and tritium mixture in microtargets of indirect irradiation (x-ray targets) were performed at the Iskra-5 facility. The density was determined according to the broadening of the lines of hydrogen- and helium-like argon added to the DT gas as a diagnostics material. A series of three experiments was performed with x-ray targets in which the central capsule filled with a DT+Ar mixture over a range of shell thicknesses. In two of the experiments, argon emission spectra were recorded and the density of the compressed gas was determined. For a microtarget approximately 280 μm in diameter with a wall approximately 7 μm thick, an analysis of the experimental results yielded an estimated density in the compressed gas of ∼1 g/cm 3 . Gas-dynamic calculations using the SNDA (spectral nonequilibrium diffusion with absorption) program show that argon emission takes place just after reaching maximum temperature, but much sooner than maximum compression. The results of a calculation for an experiment with low relative Ar concentration are in overall agreement with the experimental data. Additional investigations are needed to interpret experiments at a relatively high concentration

  2. The Jet Experiments in Nuclear Structure and Astrophysics (JENSA) gas jet target

    International Nuclear Information System (INIS)

    Chipps, K.A.; Greife, U.; Bardayan, D.W.; Blackmon, J.C.; Kontos, A.; Linhardt, L.E.; Matos, M.; Pain, S.D.; Pittman, S.T.; Sachs, A.; Schatz, H.; Schmitt, K.T.; Smith, M.S.; Thompson, P.

    2014-01-01

    New radioactive ion beam (RIB) facilities will push further away from stability and enable the next generation of nuclear physics experiments. Of great importance to the future of RIB physics are scattering, transfer, and capture reaction measurements of rare, exotic, and unstable nuclei on light targets such as hydrogen and helium. These measurements require targets that are dense, highly localized, and pure. Targets must also accommodate the use of large area silicon detector arrays, high-efficiency gamma arrays, and heavy ion detector systems to efficiently measure the reaction products. To address these issues, the Jet Experiments in Nuclear Structure and Astrophysics (JENSA) Collaboration has designed, built, and characterized a supersonic gas jet target, capable of providing gas areal densities on par with commonly used solid targets within a region of a few millimeters diameter. Densities of over 5×10 18 atoms/cm 2 of helium have been achieved, making the JENSA gas jet target the most dense helium jet achieved so far

  3. Helium turbomachinery operating experience from gas turbine power plants and test facilities

    International Nuclear Information System (INIS)

    McDonald, Colin F.

    2012-01-01

    The closed-cycle gas turbine, pioneered and deployed in Europe, is not well known in the USA. Since nuclear power plant studies currently being conducted in several countries involve the coupling of a high temperature gas-cooled nuclear reactor with a helium closed-cycle gas turbine power conversion system, the experience gained from operated helium turbomachinery is the focus of this paper. A study done as early as 1945 foresaw the use of a helium closed-cycle gas turbine coupled with a high temperature gas-cooled nuclear reactor, and some two decades later this was investigated but not implemented because of lack of technology readiness. However, the first practical use of helium as a gas turbine working fluid was recognized for cryogenic processes, and the first two small fossil-fired helium gas turbines to operate were in the USA for air liquefaction and nitrogen production facilities. In the 1970's a larger helium gas turbine plant and helium test facilities were built and operated in Germany to establish technology bases for a projected future high efficiency large nuclear gas turbine power plant concept. This review paper covers the experience gained, and the lessons learned from the operation of helium gas turbine plants and related test facilities, and puts these into perspective since over three decades have passed since they were deployed. An understanding of the many unexpected events encountered, and how the problems, some of them serious, were resolved is important to avoid them being replicated in future helium turbomachines. The valuable lessons learned in the past, in many cases the hard way, particularly from the operation in Germany of the Oberhausen II 50 MWe helium gas turbine plant, and the technical know-how gained from the formidable HHV helium turbine test facility, are viewed as being germane in the context of current helium turbomachine design work being done for future high efficiency nuclear gas turbine plant concepts. - Highlights:

  4. Experiment and analysis of basic phenomena of gas catching by liquid surface. Observation by visualizing vortex structure and gas catching process

    International Nuclear Information System (INIS)

    Kamemoto, Takashi; Nishiyama, Tadao

    1995-01-01

    Since gravity, viscous force, surface tension and so on are related simultaneously to the inertia force of flow, in gas catching phenomena, it is often difficult to grasp exactly its similarity. At the time of designing actual equipment, careful model test is required, and the validity of the evaluation by model test for applying it to actual machines sometimes becomes a problem. In this research, for the purpose of elucidating the essential mechanism of the gas-catching phenomena by vortices, and obtaining the knowledge useful for the probability of the method of evaluating the limit of gas catching, the knowledge obtained so far on the similarity law and model testing method related to the air catching by vortices was put in order, and vortex structure and basic gas-catching process were observed by water flow visualizing experiment, thus the noteworthy flow characteristics for clarifying the essential mechanism of the phenomena were obtained. The main knowledges on the air catching by vortices obtained so far, the experiment of visualizing vortices using water flow and the experimental results are reported. (K.I.)

  5. Near Detectors based on gas TPCs for neutrino long baseline experiments

    CERN Document Server

    Blondel, A

    2017-01-01

    Time Projection Chambers have been used with success for the T2K ND280 near detector and are proposed for an upgrade of the T2K near detector. High pressure TPCs are also being considered for future long-baseline experiments like Hyper-Kamiokande and DUNE. A High Pressure TPC would be a very sensitive detector for the detailed study of neutrino-nucleus interactions, a limiting factor for extracting the ultimate precision in long baseline experiments. The requirements of TPCs for neutrino detectors are quite specific. We propose here the development of state-of-the-art near detectors based on gas TPC: atmospheric pressure TPCs for T2K-II and a high-pressure TPC for neutrino experiments. The project proposed here benefits from a strong involvement of the European (CERN) members of the T2K collaboration and beyond. It is a strongly synergetic precursor of other projects of near detectors using gas TPCs that are under discussion for the long baseline neutrino projects worldwide. It will help maintain and develop...

  6. Gas-filled targets for large scalelength plasma interaction experiments on Nova

    International Nuclear Information System (INIS)

    Powers, L.V.; Berger, R.L.; Munro, D.H.

    1994-11-01

    Stimulated Brillouin backscatter from large scale length gas-filled targets has been measured on Nova. These targets were designed to approximate conditions in indirect drive ignition target designs in underdense plasma electron density (n e ∼10 21 /cm 3 ), temperature (T e >3 keV), and gradient scale lengths (L n ∼ mm, L v >6 mm) as well as calculated gain for stimulated Brillouin scattering (SBS). The targets used in these experiments were gas-filled balloons with polyimide walls (gasbags) and gas-filled hohlraums. Detailed characterization using x-ray imaging and x-ray and optical spectroscopy verifies that the calculated plasma conditions are achieved. Time-resolved SBS backscatter from these targets is <3% for conditions similar to ignition target designs

  7. Defense-waste vitrification studies during FY-1981. Summary report

    International Nuclear Information System (INIS)

    Bjorklund, W.J.

    1982-09-01

    Both simulated alkaline defense wastes and simulated acidic defense wastes (formed by treating alkaline waste with formic acid) were successfully vitrified in direct liquid-fed melter experiments. The vitrification process was improved while using the formate-treated waste. Leach resistance was essentially the same. Off-gas entrainment was the primary mechanism for material exiting the melter. When formate waste was vitrified, the flow behavior of the off gas from the melter changed dramatically from an erratic surging behavior to a more quiet, even flow. Hydrogen and CO were detectable while processing formate feed; however, levels exceeding the flamability limits in air were never approached. Two types of melter operation were tested during the year, one involving boost power. Several boosting methods located within the melter plenum were tested. When lid heating was being used, water spray cooling in the off gas was required. Countercurrent spray cooling was more effective than cocurrent spray cooling. Materials of construction for the off-gas system were examined. Inconel-690 is preferred in the plenum area. Inspection of the pilot-scale melter found that corrosion of the K-3 refractory and Inconel-690 electrodes was minimal. An overheating incident occurred with the LFCM in which glass temperatures up to 1480 0 C were experienced. Lab-scale vitrification tests to study mercury behavior were also completed this year. 53 figures, 63 tables

  8. Experience in producing LEU fuel elements for the RSG-GAS

    International Nuclear Information System (INIS)

    Suripto, A.; Soentono, S.

    1991-01-01

    To achieve a self-reliance in the operation of the 30 MW Multipurpose Research Reactor at Serpong (the RSG-GAS), a fuel element production facility has been constructed nearby. The main task of the facility is to produce MTR type fuel and control elements containing U 3 O 8 -Al dispersion LEU fuel for the RSG-GAS. The hot commissioning activity has started in early 1988 after completion of the cold commissioning using depleted uranium in 1987, marking the beginning of the real production activity. This paper briefly describes the main features of the fuel production facility, the production experience gained so far, and its current production activity. (orig.)

  9. Resistive transition for two-dimensional superconductors: Comparison between experiments and Coulomb-gas-model predictions

    International Nuclear Information System (INIS)

    Minnhagen, P.

    1983-01-01

    The Coulomb-gas model of vortex fluctuations leads to scaling relations for the resistive transition which can be directly tested by experiments. By analyzing published resistance data, it is shown that there is experimental evidence for the Coulomb-gas scaling relation in the absence of a perpendicular magnetic field. It is also shown that there exists some suggestive support for the Coulomb-gas predictions in the presence of a magnetic field

  10. Analysis of the ATLAS Cold Leg Top-Slot Break Experiment Using the MARS Code

    Energy Technology Data Exchange (ETDEWEB)

    Ha, T. W.; Jeong, J. J. [Pusan National University, Busan (Korea, Republic of)

    2016-10-15

    During a small-break loss of coolant accident (SBLOCA) or intermediate-break loss of coolant accident (IBLOCA) in a PWR, such as the APR1400, the steam volume in the reactor vessel upper plenum may continue to expand until the liquid phase in the horizontal intermediate legs is released, called loop seal clearing (LSC), due to the increase of the pressure in the upper plenum. A domestic standard problem (DSP) exercise using the ATLAS facility was promoted in order to transfer the database to domestic nuclear industries. For 4th DSP (DSP-04), the ATLAS cold leg top-slot break experiment was postulated. For the DSP-04, main concerns are to predict the LSC and LSR having a significantly effect on the behavior of the system under long term cooling. In this study, we simulated the ATLAS cold leg top-slot break experiment using the MARS code and the predicted LSC and LSR are compared to experimental results. The LTC-CL-04R was simulated using the MARS code. Most of the predicted results agree well with the experimental data. However, the timing of LSC and LSR is slightly different from each other and, thus, the behavior of the primary system is slightly different. The core heat up was not observed in the experiment and the calculation.

  11. Experiment data report for Semiscale Mod-1 Test S-05-3 (alternate ECC injection test)

    International Nuclear Information System (INIS)

    Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

    1977-03-01

    Recorded test data are presented for Test S-05-3 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-3 was conducted from initial conditions of 2263 psia and 545 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg sides of the intact and broken loops and into the vessel upper plenum to simulate emergency core coolant injection in a PWR. For Test S-05-3, specifically the effects of upper plenum coolant injection on core thermal and system response were being investigated

  12. Three-dimensional Simulation of Gas Conductance Measurement Experiments on Alcator C-Mod

    International Nuclear Information System (INIS)

    Stotler, D.P.; LaBombard, B.

    2004-01-01

    Three-dimensional Monte Carlo neutral transport simulations of gas flow through the Alcator C-Mod subdivertor yield conductances comparable to those found in dedicated experiments. All are significantly smaller than the conductance found with the previously used axisymmetric geometry. A benchmarking exercise of the code against known conductance values for gas flow through a simple pipe provides a physical basis for interpreting the comparison of the three-dimensional and experimental C-Mod conductances

  13. Numerical analysis of gas puff modulation experiment on JT-60U

    International Nuclear Information System (INIS)

    Nagashima, Keisuke; Sakasai, Akira

    1992-03-01

    In tokamak transport physics, source modulation experiments are one of the most effective methods. For an analysis of these modulation experiments, a simple numerical method was developed to solve the general transport equations. This method was applied to gas puff modulation experiments on JT-60U. From the comparison between the measured and calculated density perturbations, it was found that the particle diffusion coefficient is about 0.8 m 2 /sec in the edge region and 0.1-0.2 m 2 /sec in the central region. (author)

  14. The unique field experiments on the assessment of accident consequences at industrial enterprises of gas-chemical complexes

    International Nuclear Information System (INIS)

    Belov, N.S.; Trebin, I.S.; Sorokovikova, O.

    1998-01-01

    Sour natural gas fields are the unique raw material base for setting up such large enterprises as gas chemical complexes. The presence of high toxic H 2 S in natural gas results in widening a range of dangerous and harmful factors for biosphere. Emission of such gases into atmosphere during accidents at gas wells and gas pipelines is of especial danger for environment and first of all for people. Development of mathematical forecast models for assessment of accidents progression and consequences is one of the main elements of works on safety analysis and risk assessment. The critical step in development of such models is their validation using the experimental material. Full-scale experiments have been conducted by the All-Union Scientific-Research institute of Natural Gases and Gas Technology (VNIIGAZ) for grounding of sizes of hazard zones in case of the severe accidents with the gas pipelines. The source of emergency gas release was the working gas pipelines with 100 mm dia. And 110 km length. This pipeline was used for transportation of natural gas with significant amount of hydrogen sulphide. During these experiments significant quantities of the gas including H 2 S were released into the atmosphere and then concentrations of gas and H 2 S were measured in the accident region. The results of these experiments are used for validation of atmospheric dispersion models including the new Lagrangian trace stochastic model that takes into account a wide range of meteorological factors. This model was developed as a part of computer system for decision-making support in case of accident release of toxic gases into atmosphere at the enterprises of Russian gas industry. (authors)

  15. Experiments to Evaluate and Implement Passive Tracer Gas Methods to Measure Ventilation Rates in Homes

    Energy Technology Data Exchange (ETDEWEB)

    Lunden, Melissa [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Faulkner, David [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Heredia, Elizabeth [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Cohn, Sebastian [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Dickerhoff, Darryl [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Noris, Federico [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Logue, Jennifer [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Hotchi, Toshifumi [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Singer, Brett [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Sherman, Max H. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2012-10-01

    This report documents experiments performed in three homes to assess the methodology used to determine air exchange rates using passive tracer techniques. The experiments used four different tracer gases emitted simultaneously but implemented with different spatial coverage in the home. Two different tracer gas sampling methods were used. The results characterize the factors of the execution and analysis of the passive tracer technique that affect the uncertainty in the calculated air exchange rates. These factors include uncertainties in tracer gas emission rates, differences in measured concentrations for different tracer gases, temporal and spatial variability of the concentrations, the comparison between different gas sampling methods, and the effect of different ventilation conditions.

  16. Experiment data report for LOFT nonnuclear test L1-3

    International Nuclear Information System (INIS)

    Millar, G.M.

    1977-04-01

    Test L1-3 was the third in a series of five nonnuclear isothermal blowdown tests conducted by the Loss of Fluid Test (LOFT) Program. For this test the LOFT Facility was configured to simulate a loss-of-coolant accident in a large pressurized water reactor resulting from a 200 percent double-ended shear break in a cold leg of the primary coolant system. A hydraulic core simulator assembly was installed in place of the nuclear core. The initial conditions in the primary coolant system intact loop were: temperature at 540 0 F, pressure at 2256 psig, and loop flow at 2.34 x 10 6 lbm/hr. During system depressurization, emergency core cooling water was specified to be injected into the lower plenum of the reactor vessel using an accumulator, a low-pressure injection system pump, and a high-pressure injection system pump to provide data on the effects of emergency core cooling on the system thermal-hydraulic response. Injection into the lower plenum was initiated from the high- and low-pressure injection systems. Injection from the accumulator, however, was not initiated because a valve was inadvertently left closed. The experiment, therefore, was not completely successful in that one of the objectives outlined in the experiment operating specification for this test was not accomplished. Test L1-3 was repeated at Test L1-3A to meet the experimental requirements. Despite these difficulties, Test L1-3 did provide very valuable data to verify experiment repeatability

  17. Experiments with background gas in a vacuum arc centrifuge

    International Nuclear Information System (INIS)

    Dallaqua, R.S.; Simpson, S.W.; Del Bosco, E.

    1996-01-01

    Since promising isotope separation results were first reported by Krishnan et al. in 1981, a range of vacuum arc centrifuge experiments have been conducted in laboratories around the world. The PCEN (Plasma CENtrifuge) vacuum arc centrifuge at the Brazilian National Institute for Space Research has been used for isotope separation studies with cathode materials of carbon and magnesium and also to investigate the performance in terms of the rotational velocity attained for different cathode materials. Here, a vacuum arc centrifuge has been operated with an initial filling gas of either argon or hydrogen for pressures ranging from 10 -3 to 10 -1 Pa. The angular velocity ω of the plasma has been determined by cross-correlating the signals from potential probes, and the electron temperature T has been deduced from Langmuir probe data. At high gas pressures and early times during the 14 ms plasma lifetime, high-frequency nonuniformities frequently observed in the vacuum discharge disappear, suggesting that the associated instability is suppressed. Under the same conditions, nonuniformities rotating with much lower angular velocities are observed in the plasma. Temperatures are reduced in the presence of the background gas, and the theoretical figure of merit for separation proportional to ω 2 /T is increased compared to its value in the vacuum discharge for both argon and hydrogen gas fillings

  18. Enthalpy of Vaporization by Gas Chromatography: A Physical Chemistry Experiment

    Science.gov (United States)

    Ellison, Herbert R.

    2005-01-01

    An experiment is conducted to measure the enthalpy of vaporization of volatile compounds like methylene chloride, carbon tetrachloride, and others by using gas chromatography. This physical property was measured using a very tiny quantity of sample revealing that it is possible to measure the enthalpies of two or more compounds at the same time.

  19. Automation of experiments at Dubna Gas-Filled Recoil Separator

    Science.gov (United States)

    Tsyganov, Yu. S.

    2016-01-01

    Approaches to solving the problems of automation of basic processes in long-term experiments in heavy ion beams of the Dubna Gas-Filled Recoil Separator (DGFRS) facility are considered. Approaches in the field of spectrometry, both of rare α decays of superheavy nuclei and those for constructing monitoring systems to provide accident-free experiment running with highly radioactive targets and recording basic parameters of experiment, are described. The specific features of Double Side Silicon Strip Detectors (DSSSDs) are considered, special attention is paid to the role of boundary effects of neighboring p-n transitions in the "active correlations" method. An example of an off-beam experiment attempting to observe Zeno effect is briefly considered. Basic examples for nuclear reactions of complete fusion at 48Ca ion beams of U-400 cyclotron (LNR, JINR) are given. A scenario of development of the "active correlations" method for the case of very high intensity beams of heavy ions at promising accelerators of LNR, JINR, is presented.

  20. Theoretical and experimental comparisons of Gamble 2 argon gas puff experiments

    International Nuclear Information System (INIS)

    Thornhill, J.W.; Young, F.C.; Whitney, K.G.; Davis, J.; Stephanakis, S.J.

    1990-01-01

    A one-dimensional radiative MHD analysis of an imploding argon gas puff plasma is performed. The calculations are set up to approximate the conditions of a series of argon gas puff experiments that were carried out on the NRL Gamble II generator. Annular gas puffs (2.5 cm diameter) are imploded with a 1.2-MA peak driving current for different initial argon mass loadings. Comparisons are made with the experimental results for implosion times, K, L-shell x-ray emission, and energy coupled from the generator to the plasma load. The purpose of these calculations is to provide a foundation from which a variety of physical phenomena which influence the power and total energy of the x-ray emission can be analyzed. Comparisons with similar experimental and theoretical results for aluminum plasmas are discussed

  1. The dynamic characteristics of HTGR (High Temperature Gas Cooled Reactor) system, (2)

    International Nuclear Information System (INIS)

    Kudo, Kazuhiko; Ohta, Masao; Kawasaki, Hidenori

    1979-01-01

    The dynamic characteristics of a HTGR plant, which has two cooling loops, was investigated. The analytical model consists of the core with fuel sleeves, coolant channels and blocks, the upper and lower reflectors, the high and low temperature plenums, two double wall pipings, two intermediate heat exchangers and the secondary system. The key plant parameters for calculation were as follows: the core outlet gas temperature 1000 deg C, the reactor thermal output 50 MW, the flow rate of primary coolant gas 7.96 kg/sec-loop and the pressure of primary coolant gas 40 kg/cm 2 at the rated operating condition. The calculating parameters were fixed as follows: the time interval for core characteristic analysis 0.1 sec, the time interval for thermal characteristic analysis 5.0 sec, the number of division of fuel channels 130, and the number of division of an intermediate heat exchanger 200. The assumptions for making the model were evaluated especially for the power distribution in the core and the heat transmission coefficients in the core, the double wall piping and the intermediate heat exchangers. Concerning the analytical results, the self-control to the outer disturbance of reactivity and the plant dynamic behavior due to the change of flow rate of primary and secondary coolants, and the change of gas temperature of secondary coolant at the inlet of intermediate heat exchangers, are presented. (Nakai, Y.)

  2. Maximum total organic carbon limit for DWPF melter feed

    International Nuclear Information System (INIS)

    Choi, A.S.

    1995-01-01

    DWPF recently decided to control the potential flammability of melter off-gas by limiting the total carbon content in the melter feed and maintaining adequate conditions for combustion in the melter plenum. With this new strategy, all the LFL analyzers and associated interlocks and alarms were removed from both the primary and backup melter off-gas systems. Subsequently, D. Iverson of DWPF- T ampersand E requested that SRTC determine the maximum allowable total organic carbon (TOC) content in the melter feed which can be implemented as part of the Process Requirements for melter feed preparation (PR-S04). The maximum TOC limit thus determined in this study was about 24,000 ppm on an aqueous slurry basis. At the TOC levels below this, the peak concentration of combustible components in the quenched off-gas will not exceed 60 percent of the LFL during off-gas surges of magnitudes up to three times nominal, provided that the melter plenum temperature and the air purge rate to the BUFC are monitored and controlled above 650 degrees C and 220 lb/hr, respectively. Appropriate interlocks should discontinue the feeding when one or both of these conditions are not met. Both the magnitude and duration of an off-gas surge have a major impact on the maximum TOC limit, since they directly affect the melter plenum temperature and combustion. Although the data obtained during recent DWPF melter startup tests showed that the peak magnitude of a surge can be greater than three times nominal, the observed duration was considerably shorter, on the order of several seconds. The long surge duration assumed in this study has a greater impact on the plenum temperature than the peak magnitude, thus making the maximum TOC estimate conservative. Two models were used to make the necessary calculations to determine the TOC limit

  3. Quantum state-resolved, bulk gas energetics: Comparison of theory and experiment

    Energy Technology Data Exchange (ETDEWEB)

    McCaffery, Anthony J., E-mail: A.J.McCaffery@sussex.ac.uk [Department of Chemistry, University of Sussex, Brighton, Sussex BN1 6SJ (United Kingdom)

    2016-05-21

    Until very recently, the computational model of state-to-state energy transfer in large gas mixtures, introduced by the author and co-workers, has had little experimental data with which to assess the accuracy of its predictions. In a novel experiment, Alghazi et al. [Chem. Phys. 448, 76 (2015)] followed the equilibration of highly vibrationally excited CsH(D) in baths of H{sub 2}(D{sub 2}) with simultaneous time- and quantum state-resolution. Modal temperatures of vibration, rotation, and translation for CsH(D) were obtained and presented as a function of pump-probe delay time. Here the data from this study are used as a test of the accuracy of the computational method, and in addition, the consequent changes in bath gas modal temperatures, not obtainable in the experiment, are predicted. Despite large discrepancies between initial CsH(D) vibrational states in the experiment and those available using the computational model, the quality of agreement is sufficient to conclude that the model’s predictions constitute at least a very good representation of the overall equilibration that, for some measurements, is very accurate.

  4. City gates maintenance - TBG experience; Experiencia da TBG na manutencao de estacoes de entrega de gas natural

    Energy Technology Data Exchange (ETDEWEB)

    Rocha, Adir de Brito; Tavares, Cipriano Homem; Pinto, Jose Eduardo Christovao [Transportadora Brasileira Gasoduto Bolivia-Brasil, S.A., Rio de Janeiro, RJ (Brazil)

    2005-07-01

    Owner and sole operator in Brasilian territory of the Bolivia-Brazil Gas Pipeline (GASBOL), the largest of its kind in South America, TBG started operations on July 1 st, 1999. Since then, it has ensured transportation of Bolivian natural gas into Brazil swiftly and safely. This paper relates the TBG experiences on Natural Gas City Gates maintenance and its components like filtering, heating, pressure reducing redundant valves, turbine meters, flow computers, solar panel power supply, satellite monitoring system, acquired by 6 years of operation of Brazil Bolivia Gas Pipe Line. It describes the definitions of maintenance plans using RCM - Reliability-centered Maintenance concepts and the most important learning experiences. (author)

  5. Large-scale Experiment for Water and Gas Transport in Cementitious Backfill Materials (Phase 1 ): COLEX I

    International Nuclear Information System (INIS)

    Mayer, G.; Wittmann, F.H.; Moetsch, H.A.

    1998-05-01

    In the planned Swiss repository for low- and intermediate-level radioactive waste, the voids between the waste containers will be backfilled with a highly permeable mortar (NAGRA designation: mortar M1 ). As well as providing mechanical stability through filling of voids and sorbing radionuclides, the mortar must divert gases formed in the repository as a result of corrosion into the neighbouring host rock. This will prevent damage which could be caused by excess pressure on the repository structures. Water transport, which is coupled to gas transport, is also of interest. The former is responsible for the migration of radionuclides. Up till now, numerical simulations for a repository situation were carried out using transport parameters determined for small samples in the laboratory. However, the numerical simulations still had to be validated by a large-scale experiment. The investigations presented here should close this gap. Investigations into gas and water transport were carried out using a column (up to 5.4 m high) filled with backfill mortar. The column has a modular construction and can be sealed at the top end with a material of defined permeability (plug or top plug). The possibility to vary the material of the plug allows the influence of the more impermeable cavern lining or possible gas escape vents in the cavern roof to be investigated. A gas supply is connected to the bottom end and is used to simulate different gas generation rates from the waste. A total of 5 experiments were carried out in which the gas generation rate, the column height and the permeability of the plug were varied. Before the start of the experiments, the mortar in the column and the plug were saturated with water to approx. 95 %. In all the experiments, an increase in pressure with time could be observed. The higher the gas generation rate and the lower the permeability of the plug, the more quickly this occurred. At the beginning, only water flow out of the top of the column

  6. Small angle X-ray scattering experiments with three-dimensional imaging gas detectors

    International Nuclear Information System (INIS)

    La Monaca, A.; Iannuzzi, M.; Messi, R.

    1985-01-01

    Measurements of small angle X-ray scattering of lupolen - R, dry collagen and dry cornea are presented. The experiments have been performed with synchrotron radiation and a new three-dimensional imaging drif-chamber gas detector

  7. An improved model of fission gas atom transport in irradiated uranium dioxide

    Science.gov (United States)

    Shea, J. H.

    2018-04-01

    The hitherto standard approach to predicting fission gas release has been a pure diffusion gas atom transport model based upon Fick's law. An additional mechanism has subsequently been identified from experimental data at high burnup and has been summarised in an empirical model that is considered to embody a so-called fuel matrix 'saturation' phenomenon whereby the fuel matrix has become saturated with fission gas so that the continued addition of extra fission gas atoms results in their expulsion from the fuel matrix into the fuel rod plenum. The present paper proposes a different approach by constructing an enhanced fission gas transport law consisting of two components: 1) Fick's law and 2) a so-called drift term. The new transport law can be shown to be effectively identical in its predictions to the 'saturation' approach and is more readily physically justifiable. The method introduces a generalisation of the standard diffusion equation which is dubbed the Drift Diffusion Equation. According to the magnitude of a dimensionless Péclet number, P, the new equation can vary from pure diffusion to pure drift, which latter represents a collective motion of the fission gas atoms through the fuel matrix at a translational velocity. Comparison is made between the saturation and enhanced transport approaches. Because of its dependence on P, the Drift Diffusion Equation is shown to be more effective at managing the transition from one type of limiting transport phenomenon to the other. Thus it can adapt appropriately according to the reactor operation.

  8. NACOWA experiments on LMFBR cover gas aerosols, heat transfer, and fission product enrichment

    International Nuclear Information System (INIS)

    Minges, J.; Schuetz, W.

    1993-12-01

    Fifteen different NACOWA test series were carried out. The following items were investigated: sodium mass concentration in the cover gas, sodium aerosol particle size, radiative heat transfer across the cover gas, total heat transfer across the cover gas, sodium deposition on the cover plate, temperature profiles across the cover gas, phenomena if the argon cover gas is replaced by helium, enrichment of cesium, iodine, and zinc in the aerosol and in the deposits. The conditions were mainly related to the design parameters of the EFR. According to the first consistent design, a pool temperature of 545 C and a roof temperature of only 120 C were foreseen at a cover gas height of 85 cm. The experiments were carried out in a stainless steel test vessel of 0.6 m diameter and 1.14 m height. Pool temperature (up to 545 C), cover gas height (12.5 cm, 33 cm, and others), and roof temperature (from 110 C to 450 C) were the main parameters. (orig./HP) [de

  9. Time series analysis of pressure fluctuation in gas-solid fluidized beds

    Directory of Open Access Journals (Sweden)

    C. Alberto S. Felipe

    2004-09-01

    Full Text Available The purpose of the present work was to study the differentiation of states of typical fluidization (single bubble, multiple bubble and slugging in a gas-solid fluidized bed, using spectral analysis of pressure fluctuation time series. The effects of the method of measuring (differential and absolute pressure fluctuations and the axial position of the probes in the fluidization column on the identification of each of the regimes studied were evaluated. Fast Fourier Transform (FFT was the mathematic tool used to analysing the data of pressure fluctuations, which expresses the behavior of a time series in the frequency domain. Results indicated that the plenum chamber was a place for reliable measurement and that care should be taken in measurement in the dense phase. The method allowed fluid dynamic regimes to be differentiated by their dominant frequency characteristics.

  10. Fuel assembly

    International Nuclear Information System (INIS)

    Yamazaki, Hajime.

    1995-01-01

    In a fuel assembly having fuel rods of different length, fuel pellets of mixed oxides of uranium and plutonium are loaded to a short fuel rod. The volume ratio of a pellet-loaded portion to a plenum portion of the short fuel rod is made greater than the volume ratio of a fuel rod to which uranium fuel pellets are loaded. In addition, the volume of the plenum portion of the short fuel rod is set greater depending on the plutonium content in the loaded fuel pellets. MOX fuel pellets are loaded on the short fuel rods having a greater degree of freedom relevant to the setting for the volume of the plenum portion compared with that of a long rod fuel, and the volume of the plenum portion is ensured greater depending on the plutonium content. Even if a large amount of FP gas and He gas are discharged from the MOX fuels compared with that from the uranium fuels, the internal pressure of the MOX fuel rod during operation is maintained substantially identical with that of the uranium fuel rod, so that a risk of generating excess stresses applied to the fuel cladding tubes and rupture of fuels are greatly reduced. (N.H.)

  11. Experience in oil field processing of gas and condensate at the Shatlyk deposits

    Energy Technology Data Exchange (ETDEWEB)

    Dalmatov, V.V.; Chernikov, Ye.I.; Govorun, V.P.; Turevskiy, Ye.N.

    1983-01-01

    The operation of installations for preparing gas are analyzed, along with the operation of individual technological devices at the Shatlyk deposit, the basic things which hinder the support of the designed low temperature conditions are shown and recommendations for standardizing the operation of the technological installations are given. Experience in the operation of the gas preparation installations at the Shatlyk deposit is recommended for use in deposits being introduced into development.

  12. Conception and realization of optical diagnosis to characterize gas puffs in Z-Pinch experiments. Comparison between experiment and computation. Study of a new nozzle

    International Nuclear Information System (INIS)

    Barnier, J.N.

    1998-01-01

    The CEA develops research programs on plasma. A good way to generate such X-rays sources, is to realize Z-pinch experiments, so to realize the radial implosion on its axis of a conducting cylinder in a very high current. The AMBIORIX machine, allowing such experiments, calls for necessitates the use of gaseous conductors. The gas puff, coming from the nozzle, is ionised by a 2 MA current. The aim of this thesis is the characterisation of the gas source before the current impulse. For this purpose many optic diagnostics have been tested. Interferometric measures allow the gas profile density measurement. Various gas have been studied: neon, argon, helium and aluminium. For the aluminium, the resonant interferometric imagery method has been used. A new nozzle with an innovative injection technic, has been designed, characterized and tested in Z-pinch configuration. Finally measures of light diffusion (Rayleigh) have been realised to show dust in the gas. (A.L.B.)

  13. Steady-state irradiation testing of U-Pu-Zr fuel to >18% burnup

    International Nuclear Information System (INIS)

    Pahl, R.G.; Wisner, R.S.; Billone, M.C.; Hofman, G.L.

    1990-01-01

    Tests of austenitic stainless steel clad U-xP-10Zr fuel (x=o, 8, 19 wt. %) to peak burnups as high as 18.4 at. % have been completed in the EBR-II. Fuel swelling and fractional fission gas release are slowly increasing functions of burnup beyond 2 at. % burnup. Increasing plutonium content in the fuel reduces swelling and decreases the amount of fission gas which diffuses from fuel to plenum. LIFE-METAL code modelling of cladding strains is consistent with creep by fission gas loading and irradiation-induced swelling mechanisms. Fuel/cladding chemical interaction involves the ingress of rare-earth fission products. Constituent redistribution in the fuel had not limited steady-state performance. Cladding breach behavior at closure welds, in the gas plenum, and in the fuel column region have been benign events. 3 refs., 5 figs

  14. Gas reactor and associated nuclear experience in the UK relevant to high temperature reactor engineering

    International Nuclear Information System (INIS)

    Beech, D.J.; May, R.

    2000-01-01

    In the UK, the NNC played a leading role in the design and build of all of the UK's commercial magnox reactors and advanced gas-cooled reactors (AGRs). It was also involved in the DRAGON project and was responsible for producing designs for large scale HTRs and other gas reactor designs employing helium and carbon dioxide coolants. This paper addresses the gas reactor experience and its relevance to the current HTR designs under development which use helium as the coolant, through the consideration of a representative sample of the issues addressed in the UK by the NNC in support of the AGR and other reactor programmes. Modern HTR designs provide unique engineering challenges. The success of the AGR design, reflected in the extended lifetimes agreed upon by the licensing authorities at many stations, indicates that these challenges can be successfully overcome. The UK experience is unique and provides substantial support to future gas reactor and high temperature engineering studies. (authors)

  15. Developments for transactinide chemistry experiments behind the gas-filled separator TASCA

    International Nuclear Information System (INIS)

    Even, Julia

    2011-01-01

    Topic of this thesis is the development of experiments behind the gas-filled separator TASCA (TransActinide Separator and Chemistry Apparatus) to study the chemical properties of the transactinide elements. In the first part of the thesis, the electrodepositions of short-lived isotopes of ruthenium and osmium on gold electrodes were studied as model experiments for hassium. From literature it is known that the deposition potential of single atoms differs significantly from the potential predicted by the Nernst equation. This shift of the potential depends on the adsorption enthalpy of therndeposited element on the electrode material. If the adsorption on the electrode-material is favoured over the adsorption on a surface made of the same element as the deposited atom, the electrode potential is shifted to higher potentials. This phenomenon is called underpotential deposition. Possibilities to automatize an electro chemistry experiment behind the gas-filled separator were explored for later studies with transactinide elements. The second part of this thesis is about the in-situ synthesis of transition-metal-carbonyl complexes with nuclear reaction products. Fission products of uranium-235 and californium-249 were produced at the TRIGA Mainz reactor and thermalized in a carbon-monoxide containing atmosphere. The formed volatile metal-carbonyl complexes could be transported in a gas-stream. Furthermore, short-lived isotopes of tungsten, rhenium, osmium, and iridium were synthesised at the linear accelerator UNILAC at GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt. The recoiling fusion products were separated from the primary beam and the transfer products in the gas-filled separator TASCA. The fusion products were stopped in the focal plane of TASCA in a recoil transfer chamber. This chamber contained a carbon-monoxide - helium gas mixture. The formed metal-carbonyl complexes could be transported in a gas stream to various experimental setups. All

  16. Upper plenum break LOCA investigation in the ISB-VVER and PSB-VVER facilities

    Energy Technology Data Exchange (ETDEWEB)

    Blinkov, V. N.; Melikhov, O. I.; Lipatov, I. A.; Nikonov, S. M.; Dremin, G. I.; Galchanskaya, S. A.; Gashenko, M. P.; Rovnov, A. A.; Kapustin, A. V.; Elkin, I. V. [EREC, Moscow (Russian Federation)

    2003-07-01

    The capability to define the actual NPP transient/accident scenario depends to a great extent on facilities' scaling and reliability of the system thermalhydraulic codes which, in turn, are assessed against the experimental data taken in the same facilities. At the present time, it is received fact that the rigorous modeling of the cumulative set of all thermalhydraulic processes in the plant primary and secondary sides during accident is unfeasible. Therefore, the extrapolation of the facilities loops behavior to the actual systems constitutes a fundamental problem in this area. In the paper, some aspects for the problem have been discussed in the course of comparative analysis of the data derived from the 11 % upper plenum break LOCA tests performed in the PSB-VVER and ISB-VVER integral test facilities under the close scenarios. Both facilities, PSB-VVER and ISB-VVER, are modeled the same VVER-1000 reactor in different scales. The thermalhydraulic behavior of the primary systems in both facilities has been discussed in the paper, and shown to be similar. Also, the attention has been focused upon the discrepancies in the significant variables trends. The discrepancies are shown to be caused by influence of peculiarities of the facilities hardware and due to the scale factor. The scaling study is an important aspect of the thermalhydraulic codes verification procedure. Being qualified against the experimentally simulated accident sequence in two test facilities of different scales, the thermalhydraulic codes will be capable of evaluation of the prototype behavior to greater accuracy.

  17. Noble Gas Leak Detector for Use in the SNS Neutron Electric Dipole Moment Experiment

    Science.gov (United States)

    Barrow, Chad; Huffman, Paul; Leung, Kent; Korobkina, Ekaterina; White, Christian; nEDM Collaboration Collaboration

    2017-09-01

    Common practice for leak-checking high vacuum systems uses helium as the probing gas. However, helium may permeate some materials at room temperature, making leak characterization difficult. The experiment to find a permanent electric dipole moment of the neutron (nEDM), to be conducted at Oak Ridge National Laboratories, will employ a large volume of liquid helium housed by such a helium-permeable composite material. It is desirable to construct a leak detector that can employ alternative test gases. The purpose of this experiment is to create a leak detector that can quantify the argon gas flux in a high vacuum environment and interpret this flux as a leak-rate. This apparatus will be used to check the nEDM volumes for leaks at room temperature before cooling down to cryogenic temperatures. Our leak detector uses a residual gas analyzer and a vacuum pumping station to characterize the gas present in an evacuated volume. The introduction of argon gas into the system is interpreted as a leak-rate into the volume. The device has been calibrated with NIST certified calibrated leaks and the machine's sensitivity has been calculated using background gas analysis. As a result of the device construction and software programming, we are able to leak-check composite and polyamide volumes This work was supported in part by the US Department of Energy under Grant No. DE-FG02-97ER41042.

  18. Nuclear fuel element containing particles of an alloyed Zr, Ti, and Ni getter material

    International Nuclear Information System (INIS)

    Grossman, L.N.; Levin, H.A.

    1975-01-01

    A nuclear fuel element for use in the core of a nuclear reactor is disclosed. The nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles of alloy. The container is preferably held in the spring in the plenum of the fuel element. (Official Gazette)

  19. Development of the design of the High Temperature Gas Cooled Reactor experiment

    International Nuclear Information System (INIS)

    Lockett, G.E.; Huddle, R.A.U.

    1960-01-01

    Early in 1956 a small team was formed at the Atomic Energy Research Establishment, Harwell, to investigate the possibilities of the High Temperature Gas Cooled (H.T.G.C.) Reactor System. Although the primary objective of this team was to carry out a feasibility study of the system as a whole, it soon became apparent that, in addition to design studies and economic surveys of power producing reactors, the most appropriate approach to such a novel system was to carry out a design study of a relatively small (10 to 20 M.W.) Reactor Experiment, together with the necessary research and development work associated with such a reactor. This work proceeded within the U.K.A.E.A. during the three following years, and it was felt that realistic design proposals could be put forward with sufficient confidence to justify the detailed design and construction of a 20 M.W. Reactor Experiment. In April 1959 responsibility for this Reactor Experiment was taken over by the O.E.E.C. High Temperature Gas Cooled Reactor Project, the DRAGON Project, at the Atomic Energy Establishment, Winfrith, Dorset. In this Paper the research, development, and design work is reviewed, and the proposals for the Reactor Experiment are summarised. (author)

  20. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    Energy Technology Data Exchange (ETDEWEB)

    Hastowo, Hudi; Tarigan, Alim [Multipurpose Reactor Center, National Nuclear Energy Agency of the Republic of Indonesia (PRSG-BATAN), Kawasan PUSPIPTEK Serpong, Tangerang (Indonesia)

    1999-08-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U{sub 3}O{sub 8}-Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  1. Operation experience of the Indonesian multipurpose research reactor RSG-GAS

    International Nuclear Information System (INIS)

    Hastowo, Hudi; Tarigan, Alim

    1999-01-01

    RSG-GAS is a multipurpose research reactor with nominal power of 30 MW, operated by BATAN since 1987. The reactor is an open pool type, cooled and moderated with light water, using the LEU-MTR fuel element in the form of U 3 O 8 -Al dispersion. Up to know, the reactor have been operated around 30,000 hours to serve the user. The reactor have been utilized to produce radioisotope, neutron beam experiments, irradiation of fuel element and its structural material, and reactor physics experiments. This report will explain in further detail concerning operational experience of this reactor, i.e. reactor operation data, reactor utilization, research program, technical problems and it solutions, plant modification and improvement, and development plan to enhance better reactor operation performance and its utilization. (author)

  2. Preliminary findings of the Viking gas exchange experiment and a model for Martian surface chemistry

    International Nuclear Information System (INIS)

    Oyama, V.I.; Berdahl, B.J.; Carle, G.C.

    1977-01-01

    It is stated that O 2 and CO 2 were evolved from humidified Martian soil in the gas exchange experiment on Viking Lander 1. Small changes in N 2 gas were also recorded. A model of the morphology and a hypothesis of the mechanistics of the Martian surface are proposed. (author)

  3. Post-irradiation examination of fifteen UO2/PuO2-fuel pins from the experiment DFR-350

    International Nuclear Information System (INIS)

    Geithoff, D.

    1975-06-01

    Within the framework of the fuel pin development for a sodium-cooled fast reactor a subassembly containing 77 fuel pins has been irradiated up to 5.65% fima in the Dounreay fast reactor. The pins were prototypes in terms of fuel and cladding material. The fuel consisted of mechanically mixed UO 2 (80%) and PuO 2 (20%) pressed into pellets whereas austenitic steels (W.-No. 1,4961 and 1,4988) were used as cladding material. Furthermore a blanket column of UO 2 pellets and a gas plenum were incorporated in the pin. For irradiation the conditions in a fast breeder were simulated by a linear rod power of 450 W/cm and a maximum cladding temperature of 630 0 C. After the successful completion of the irradiation, the subassembly was dismantled and fifteen pins were selected for a nondestructive and destructive examination. The tests included visual control, measurement of external dimensions, γ-spectroscopy, X-ray radiography, fission gas measurement, ceramography, radiochemical burn-up measurement. The results are presented. The most important results of the examinations seem to be the migration of fission product cesium and the fact that no signs of impending pin failure have been found. Thus the pin specification tested in this experiment is capable of achieving higher burnups under the irradiation conditions described above. (orig./AK) [de

  4. A gas microstrip wide angle X-ray detector for application in synchrotron radiation experiments

    CERN Document Server

    Bateman, J E; Derbyshire, G E; Duxbury, D M; Lipp, J; Mir, J A; Simmons, J E; Spill, E J; Stephenson, R; Dobson, B R; Farrow, R C; Helsby, W I; Mutikainen, R; Suni, I

    2002-01-01

    The Gas Microstrip Detector has counting rate capabilities several orders of magnitude higher than conventional wire proportional counters while providing the same (or better) energy resolution for X-rays. In addition the geometric flexibility provided by the lithographic process combined with the self-supporting properties of the substrate offers many exciting possibilities for X-ray detectors, particularly for the demanding experiments carried out on Synchrotron Radiation Sources. Using experience obtained in designing detectors for Particle Physics we have developed a detector for Wide Angle X-ray Scattering studies. The detector has a fan geometry which makes possible a gas detector with high detection efficiency, sub-millimetre spatial resolution and good energy resolution over a wide range of X-ray energy. The detector is described together with results of experiments carried out at the Daresbury Laboratory Synchrotron Radiation Source.

  5. Explaining experience curves for new energy technologies. A case study of liquefied natural gas

    International Nuclear Information System (INIS)

    Greaker, Mads; Lund Sagen, Eirik

    2008-01-01

    Many new energy technologies seem to experience a fall in unit price as they mature. In this paper we study the unit price of liquefying natural gas in order to make it transportable by ship to gas power installations all over the world. Our point of departure is the experience curve approach, however unlike many other studies of new energy technologies, we also seek to account for autonomous technological change, scale effects and the effects of upstream competition among technology suppliers. To our surprise we find that upstream competition is by far the most important factor contributing to the fall in unit price. With respect to the natural gas business, this may have implications for the future development in prices as the effect of increased upstream competition is temporary and likely to weaken a lot sooner than effects from learning and technological change. Another more general policy implication, is that while promoting new energy technologies, governments must not forget to pay attention to competition policy. (author)

  6. Estimation of Knudsen diffusion coefficients from tracer experiments conducted with a binary gas system and a porous medium

    Science.gov (United States)

    Hibi, Yoshihiko; Kashihara, Ayumi

    2018-03-01

    A previous study has reported that Knudsen diffusion coefficients obtained by tracer experiments conducted with a binary gas system and a porous medium are consistently smaller than those obtained by permeability experiments conducted with a single-gas system and a porous medium. To date, however, that study is the only one in which tracer experiments have been conducted with a binary gas system. Therefore, to confirm this difference in Knudsen diffusion coefficients, we used a method we had developed previously to conduct tracer experiments with a binary carbon dioxide-nitrogen gas system and five porous media with permeability coefficients ranging from 10-13 to 10-11 m2. The results showed that the Knudsen diffusion coefficient of N2 (DN2) (cm2/s) was related to the effective permeability coefficient ke (m2) as DN2 = 7.39 × 107ke0.767. Thus, the Knudsen diffusion coefficients of N2 obtained by our tracer experiments were consistently 1/27 of those obtained by permeability experiments conducted with many porous media and air by other researchers. By using an inversion simulation to fit the advection-diffusion equation to the distribution of concentrations at observation points calculated by mathematically solving the equation, we confirmed that the method used to obtain the Knudsen diffusion coefficient in this study yielded accurate values. Moreover, because the Knudsen diffusion coefficient did not differ when columns with two different lengths, 900 and 1500 mm, were used, this column property did not influence the flow of gas in the column. The equation of the dusty gas model already includes obstruction factors for Knudsen diffusion and molecular diffusion, which relate to medium heterogeneity and tortuosity and depend only on the structure of the porous medium. Furthermore, there is no need to take account of any additional correction factor for molecular diffusion except the obstruction factor because molecular diffusion is only treated in a multicomponent

  7. Multiple scattering effects in fast neutron polarization experiments using high-pressure helium-xenon gas scintillators as analyzers

    International Nuclear Information System (INIS)

    Tornow, W.; Mertens, G.

    1977-01-01

    In order to study multiple scattering effects both in the gas and particularly in the solid materials of high-pressure gas scintillators, two asymmetry experiments have been performed by scattering of 15.6 MeV polarized neutrons from helium contained in stainless steel vessels of different wall thicknesses. A monte Carlo computer code taking into account the polarization dependence of the differential scattering cross sections has been written to simulate the experiments and to calculate corrections for multiple scattering on helium, xenon and the gas containment materials. Besides the asymmetries for the various scattering processes involved, the code yields time-of-flight spectra of the scattered neutrons and pulse height spectra of the helium recoil nuclei in the gas scintillator. The agreement between experimental results and Monte Carlo calculations is satisfactory. (Auth.)

  8. LFCM [liquid-fed eramic melter] emission and off-gas system performance for feed component cesium

    International Nuclear Information System (INIS)

    Goles, R.W.; Andersen, C.M.

    1986-09-01

    Except for volatile off-gas effluents, overall adequacy of the liquid-fed ceramic melter (LFCM) system depends most upon its effectiveness in dealing with cesium. However, the mechanism responsible for melter cesium losses has proved insensitive to many LFCM operating and processing conditions. As a result, variations in inleakage, plenum temperature, feeding rate and waste loading do not significantly influence melter cesium performance. Feed composition, specifically halogen content, is the only processing variable that has had a significant effect. Due to the submicron nature of LFCM-generated aerosols, melter disengagement design features are not expected to be particularly effective in reducing cesium emission rates. For the same reason, the cesium performance of conventional quench scrubbers is quite low, being dependent only upon the magnitude of melter entrainment losses. Although a deep bed washable filter has been effective in removing submicron aerosols from the process exhaust, high performance has only been achieved under dry operating conditions. The melter's idling state does not appear to place additional demands upon the off-gas treatment system

  9. High pressure deuterium-tritium gas target vessels for muon-catalyzed fusion experiments

    International Nuclear Information System (INIS)

    Caffrey, A.J.; Spaletta, H.W.; Ware, A.G.; Zabriskie, J.M.; Hardwick, D.A.; Maltrud, H.R.; Paciotti, M.A.

    1989-01-01

    In experimental studies of muon-catalyzed fusion, the density of the hydrogen gas mixture is an important parameter. Catalysis of up to 150 fusions per muon has been observed in deuterium-tritium gas mixtures at liquid hydrogen density; at room temperature, such densities require a target gas pressure of the order of 1000 atmospheres (100 MPa, 15,000 psi). We report here the design considerations for hydrogen gas target vessels for muon-catalyzed fusion experiments that operate at 1000 and 10,000 atmospheres. The 1000 atmosphere high pressure target vessels are fabricated of Type A-286 stainless steel and lined with oxygen-free, high-conductivity (OFHC) copper to provide a barrier to hydrogen permeation of the stainless steel. The 10,000 atmosphere ultrahigh pressure target vessels are made from 18Ni (200 grade) maraging steel and are lined with OFHC copper, again to prevent hydrogen permeation of the steel. In addition to target design features, operating requirements, fabrication procedures, and secondary containment are discussed. 13 refs., 3 figs., 1 tab

  10. Process systems of PHWR - Indian experience

    Energy Technology Data Exchange (ETDEWEB)

    Ramandan, T S.V. [Madras Atomic Power Station (MAPS), Madras (India)

    1991-04-01

    Three operational problems are discussed in this paper. The reactors in Madras Atomic Power Station (MAPS), India are Pressurised Heavy Water Reactors PHWR), similar to Douglas Point PGS. The moderator heavy water is pumped into the bottom half of the calandria (horizontal reactor vessel) through one inlet manifold plenum chamber and horizontal louvers which help to distribute the moderator evenly at a very low velocity. The outlet from the calandria is through a smaller manifold structure at a higher elevation. The moderator is held on the shell side of the calandria by means of helium gas pressure differential between top of calandria and dump tank located below. The primary coolant system consists of 306 coolant channels containing the fuel and steam generators (SGs) and pumps on either side of the reactor. Each SC consists of 11 Nos. inverted U tube vertical heat exchangers where heat is transferred from primary coolant heavy water to secondary light water to produce steam. (author)

  11. Process systems of PHWR - Indian experience

    International Nuclear Information System (INIS)

    Ramandan, T.S.V.

    1991-01-01

    Three operational problems are discussed in this paper. The reactors in Madras Atomic Power Station (MAPS), India are Pressurised Heavy Water Reactors PHWR), similar to Douglas Point PGS. The moderator heavy water is pumped into the bottom half of the calandria (horizontal reactor vessel) through one inlet manifold plenum chamber and horizontal louvers which help to distribute the moderator evenly at a very low velocity. The outlet from the calandria is through a smaller manifold structure at a higher elevation. The moderator is held on the shell side of the calandria by means of helium gas pressure differential between top of calandria and dump tank located below. The primary coolant system consists of 306 coolant channels containing the fuel and steam generators (SGs) and pumps on either side of the reactor. Each SC consists of 11 Nos. inverted U tube vertical heat exchangers where heat is transferred from primary coolant heavy water to secondary light water to produce steam. (author)

  12. HM modelling of in-situ gas injection tests in bentonite and argillite: the PGZ experiment

    International Nuclear Information System (INIS)

    Gerard, P.; Charlier, R.; Radu, J.P.; La Vaissiere, R. de; Talandier, J.; Collin, F.

    2010-01-01

    Document available in extended abstract form only. During long-term repository of high and intermediate level nuclear waste in deep argillaceous geological formation, steel containers will corroded and organic material will be irradiated. The two processes lead mostly to hydrogen production. This study deals with the numerical modelling of the gas migration in both the host formation and a bentonite plug, with an emphasis on coupling between the gas transfer and the mechanical strains and stresses. More particularly the study aims to support the design of the PGZ in situ experiment that will be performed by Andra in its underground laboratory at Bure. The objective of the experiment is the analysis of the dynamics of the bentonite plug re-saturation, studying the competition between the liquid water coming from the argillite and a gas injection. The modelled experiment consists of a borehole drilled in rock clay, inside which a plug of MX-80 bentonite is set. The bentonite is naturally re-saturated by water coming from the host formation. At the same time a gas pressure, higher than the initial water pressure in the host rock, is imposed at both ends of the plug. The developed model takes into account the coupling between the mechanical behaviour and the water and gas transfers in undisturbed geo-materials. It manages explicitly liquid and vapour water, gaseous and dissolved hydrogen. Elastoplastic and non-linear elastic model are used to model the behaviour of, respectively, the argillaceous rock and the bentonite. The numerical results show the small desaturation obtained in bentonite and argillite. The influence of the coupling of the mechanic on the water and gas transfers is thus limited (due to the Bishop's effective stress). The swelling of the bentonite plug is not hindered by the gas migration and the confining effect of the engineered barrier is maintained. An analysis is made of the influence of the main transfer rock properties on the gas pressure

  13. Compaction wave profiles: Simulations of gas gun experiments

    International Nuclear Information System (INIS)

    Menikoff, Ralph

    2001-01-01

    Mesoscale simulations of a compaction wave in a granular bed of HMX have been performed. The grains are fully resolved in order that the compaction, i.e., the porosity behind the wave front, is determined by the elastic-plastic response of the grains rather than by an empirical law for the porosity as a function of pressure. Numerical wave profiles of the pressure and velocity are compared with data from a gas gun experiment. The experiment used an initial porosity of 36%, and the wave had a pressure comparable to the yield strength of the grains. The profiles are measured at the front and back of the granular bed. The transit time for the compaction wave to propagate between the gauges determines the wave speed. The wave speed depends on the porosity behind the wave and is affected by the strength model. The yield strength needed to match the experimental wave speed is discussed. Analysis of the lead wave through the granular bed, based on impedance matches using the Hugoniot loci, indicates that the compaction wave triggers a small amount of burn, less than 1% mass fraction, on the microsecond time scale of the experiment. copyright 2001 American Institute of Physics

  14. Gas Chromatographic Determination of Methyl Salicylate in Rubbing Alcohol: An Experiment Employing Standard Addition.

    Science.gov (United States)

    Van Atta, Robert E.; Van Atta, R. Lewis

    1980-01-01

    Provides a gas chromatography experiment that exercises the quantitative technique of standard addition to the analysis for a minor component, methyl salicylate, in a commercial product, "wintergreen rubbing alcohol." (CS)

  15. Unsteady Reynolds Averaged Navier-Stokes and Large Eddy Simulations of Flows across Staggered Tube Bundle for a VHTR Lower Plenum Design

    International Nuclear Information System (INIS)

    Choi, Hyeon Kyeong; Park, Jong Woon

    2013-01-01

    In this work, behavior of unsteady and oscillating flow through a typical tube bundle array are analyzed by unsteady computations: 2D unsteady Reynolds averaged Navier-Stokes (URANS) and 3D Large Eddy Simulation (LES) and the results are compared with existing experimental data. In order to confirm appropriateness and limitations of CFD applications in the Korean VHTR design, two types of unsteady computations are performed such as 2D unsteady Reynolds averaged Navier-Stokes (URANS) and 3D Large Eddy Simulation (LES) for the existing tube bundle array. The velocity component profiles are compared with the experimental data and it is concluded that the URANS with the standard k-ω model is reasonably appropriate for cost-effective VHTR lower plenum analysis. Nevertheless, if more accurate results are needed, the LES-Smagorinsky computation is recommended considering limitations in the time averaged RANS in capturing small eddies

  16. Operating experiences of gas purification system of Heavy Water Plant Talcher (Paper No. 1.11)

    International Nuclear Information System (INIS)

    Bhattacharya, R.; Mohanty, P.R.; Pandey, B.L.

    1992-01-01

    The operating experiences with the purification system installed at Heavy Water Plant, Talcher for purification of feed synthesis gas from fertilizer plant is described. The purification system has performed satisfactorily even with levels of impurities as much as 15 to 20 ppm of oxygen and carbon monoxide. The system could not however be tested at designed gas throughput and on a sustained basis. However, increase in gas throughput upto the design value is not expected to pose any problem on the performance of the purification system. (author). 5 figs

  17. Cover gas purification experience at KNK

    Energy Technology Data Exchange (ETDEWEB)

    Richard, H; Stade, K Ch [Kernkraftwerk-Betriebsgesellschaft m.b.H., Eggenstein-Leopoldshafen (Germany); Stamm, H H [Institute of Radiochemistry, Nuclear Research Center, Karsruhe (Germany)

    1987-07-01

    KNK II is an experimental, sodium cooled fast breeder reactor. The reactor was operated until 1974 with a thermal core (KNK I). The plant was converted into a fast breeder reactor (KNK II) from 1974 to 1977. The commissioning of KNK II was started in October 1977 with the first fast core KNK 11/1. After 400 effective full power days (EFPD) the reactor was shut down in August 1982. After replacing the complete core by the second fast core KNK 11/2, the plant went into operation again in August 1983. In August 1986 nearly 400 EFPD were achieved with the second fast core. Argon is used as cover gas in the primary and secondary sodium systems of KNK. In former times fresh argon was supplied by a bundle of gas cylinders. Later on a liquid argon supply was installed. Purification of cover gas is done by flushing only. During KNK I operation no fuel failures occurred. The primary cover gas activity was characterized by the formation of Ar-41, only small quantities of fission gas were measured, released from 'tramp uranium'. Therefore, no problems existed during KNK I operation with regard to radioactive gas disposal. However, after start-up of KNK II, several fuel elements failed. Until August 1986, five fuel failures were observed, two in KNK 11/1, and three in KNK 11/2. Sometimes, operation with defective fuel pins caused problems when fission gases leaked into the containment atmosphere, and the access had to be restricted. The purging rate of the primary cover gas was limited by the capacity of the charcoal filters in the delay line. Of all non-radioactive impurities, hydrogen (H{sub z}) and nitrogen (N{sub 2}) were of most importance in the primary cover gas. Main source of both impurities was the ingress of air and atmospheric moisture during handling operations in shutdown periods. An other possible source for hydrogen might be a release from the steel-clad zirconium hydride, used as moderator in the moderated driver fuel elements. Additional nitrogen may diffuse

  18. Forty years of experience on closed-cycle gas turbines

    International Nuclear Information System (INIS)

    Keller, C.

    1978-01-01

    Forty years of experience on closed-cycle gas turbines (CCGT) is emphasized to substantiate the claim that this prime-mover technology is well established. European fossil-fired plants with air as the working fluid have been individually operated over 100,000 hours, have demonstrated very high availability and reliability, and have been economically successful. Following the initial success of the small air closed cycle gas turbine plants, the next step was the exploitation of helium as the working fluid for plants above 50 MWe. The first fossil fired combined power and heat plant at Oberhausen, using a helium turbine, plays an important role for future nuclear systems and this is briefly discussed. The combining of an HTGR and an advanced proven power conversion system (CCGT) represents the most interesting and challenging project. The key to acceptance of the CCGT in the near term is the introduction of a small nuclear cogeneration plant (100 to 300 MWe) that utilizes the waste heat, demonstrating a very high fuel utilization efficiency: aspects of such a plant are outlined. (author)

  19. Identification of gap cooling phenomena from LAVA-4 experiment using MELCOR

    International Nuclear Information System (INIS)

    Park, Jong-Hwa; Kim, Dong-Ha; Kim, See-Darl; Kim, Sang-Baik; Kim, Hee-Dong

    2000-01-01

    During the severe accident, whether the hot debris in. lower head will be cool-down or not is the important issue concerning the plant safety. KAERI has launched the 'LAVA' experimental program to examine the existence of initial gap and its effect on the cooling of hot debris. The objective of this study is to identify the gap cooling phenomena from the analysis of simulation results on LAVA-4 experiment using MELCOR1.8.4 code. Three parameters on the debris coolability in MELCOR are the quenching heat transfer coefficient for the interaction between molten Al 2 O 3 and water, the heat transfer coefficient from debris to wall and the diameter of the particulate debris for calculating the available heat transfer area with water. The sensitivity study was performed with these three parameters. However it was believed that there must be a gap between debris and inside wall during the transient. MELCOR1.8.4 does not consider these gap-cooling phenomena. Therefore a conceptual gap-cooling model has been developed and implemented into the lower plenum model in MELCOR to take into account the gap effect in the lower plenum. When the 'gap model' is implemented, the peak temperature of the vessel wall was reduced and its cooling rate was increased. (author)

  20. Preliminary results of Resistive Plate Chambers operated with eco-friendly gas mixtures for application in the CMS experiment

    International Nuclear Information System (INIS)

    Abbrescia, M.; Muhammad, S.; Saviano, G.; Auwegem, P. Van; Cauwenbergh, S.; Tytgat, M.; Benussi, L.; Bianco, S.; Passamonti, L.; Pierluigi, D.; Piccolo, D.; Primavera, F.; Russo, A.; Ferrini, M.

    2016-01-01

    The operations of Resistive Plate Chambers in LHC experiments require Fluorine based (F-based) gases for optimal performance. Recent European regulations demand the use of environmentally unfriendly F-based gases to be limited or banned. In view of the CMS experiment upgrade, several tests are ongoing to measure the performance of the detector with these new ecological gas mixtures, in terms of efficiency, streamer probability, induced charge and time resolution. Prototype chambers with readout pads and with the standard CMS electronic setup are under test. In this paper preliminary results on performance of RPCs operated with a potential eco-friendly gas candidate 1,3,3,3-Tetrafluoropropene, commercially known as HFO-1234ze, with CO 2 and CF 3 I based gas mixtures are presented and discussed for the possible application in the CMS experiment.

  1. Preliminary results of Resistive Plate Chambers operated with eco-friendly gas mixtures for application in the CMS experiment

    CERN Document Server

    Abbrescia, M.

    2016-01-01

    The operations of Resistive Plate Chambers in LHC experiments require Fluorine based (F-based) gases for optimal performance. Recent European regulations demand the use of environmentally unfriendly F-based gases to be limited or banned. In view of the CMS experiment upgrade, several tests are ongoing to measure the performance of the detector with these new ecological gas mixtures, in terms of efficiency, streamer probability, induced charge and time resolution. Prototype chambers with readout pads and with the standard CMS electronic setup are under test. In this paper preliminary results on performance of RPCs operated with a potential eco-friendly gas candidate 1,3,3,3-Tetrafluoropropene, commercially known as HFO-1234ze, with CO2 and CF3I based gas mixtures are presented and discussed for the possible application in the CMS experiment.

  2. Commissioning experiment of the polarized internal gas target with deuterium at ANKE/COSY

    Energy Technology Data Exchange (ETDEWEB)

    Gou, Boxing [Institut fuer Kernphysik, Forschungszentrum Juelich (Germany); Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou (China); Collaboration: ANKE-Collaboration

    2012-07-01

    In order to conduct the production experiments with polarized deuterium target and (un)polarized proton beam at ANKE/COSY, a commissioning experiment of the polarized internal target with deuterium is imperative. The commissioning experiment includes the measurements of both the vector (Q{sub y}) and tensor (Q{sub yy}) polarization of the deuterium gas target through the nuclear reactions with large and well known analyzing powers, which can be detected in ANKE. The dependence of the polarizations along the storage cell is also determined. The poster presents the physics case for the experiments with deuterium polarized internal target and the apparatus needed for the commissioning experiment, as well as the procedure of extraction for spin observables.

  3. Nondestructive post-irradiation examination of Loop-1, S1 and B1 rods

    International Nuclear Information System (INIS)

    Bratton, R.L.

    1997-05-01

    As a part of the Pacific Northwest National Laboratory's Tritium Target Development Program, eleven tritium target rods were irradiated in the Advanced Test Reactor located at the Idaho National Engineering and Environmental Laboratory during 1991. Both nondestructive and destructive post-irradiation examination on all eleven rods was planned under the Tritium Target Development Program. Funding for the program was reduced in 1991 resulting in the early removal of the program experiments before reaching their irradiation goals. Post-irradiation examination was only performed on one of the irradiated rods at the Pacific Northwest National Laboratory before the program was terminated in 1992. On December 6, 1995, the Secretary of Energy announced the pursuit of the Commercial Light-Water Reactor option for producing tritium establishing the Tritium Target Qualification Program at the Pacific Northwest National Laboratory. This program decided to pursue nondestructive and destructive post-irradiation examination of the ten remaining rods from the previous program. The ten rods comprise three experiments. The Loop-1 experiment irradiated eight target rods in a loop configuration for 217 irradiation days. The other two rods were irradiated in two separate irradiation experiments, designated as S1 and B1 for 143 effective full-power days, but at different power levels. After the ten rods were transferred from the ATR Canal to the Hot Fuels Examination Facility, the following examinations were performed: (1) visual examination and photography; (2) neutron radiography; (3) axial gamma scanning; (4) contact profilometry measurement; (5) bow and length measurements; (6) rod puncture and plenum gas analysis/measurement of plenum gas quantity; (7) void volume determination; and (8) internal pressure determination. This report presents the data collected during these examinations

  4. CELSS experiment model and design concept of gas recycle system

    Science.gov (United States)

    Nitta, K.; Oguchi, M.; Kanda, S.

    1986-01-01

    In order to prolong the duration of manned missions around the Earth and to expand the human existing region from the Earth to other planets such as a Lunar Base or a manned Mars flight mission, the controlled ecological life support system (CELSS) becomes an essential factor of the future technology to be developed through utilization of space station. The preliminary system engineering and integration efforts regarding CELSS have been carried out by the Japanese CELSS concept study group for clarifying the feasibility of hardware development for Space station experiments and for getting the time phased mission sets after FY 1992. The results of these studies are briefly summarized and the design and utilization methods of a Gas Recycle System for CELSS experiments are discussed.

  5. Summarizing evaluation of the results of in-pile experiments for the transient fission gas release under accidental conditions of fast breeders

    International Nuclear Information System (INIS)

    Fischer, E.A.; Vaeth, L.

    1989-04-01

    The transient fission gas behaviour and the fission gas induced fuel motion were studied in in-pile experiments in different countries, under conditions typical for hypothetical accidents. This report summarizes first the different experiment series and the main results. Then, a comparative evaluation is given, which provides a basis for the choice of the fission gas parameters in the accident code SAS3D

  6. Energy derivatives, experiences in derivative trading from the natural gas industry : back to basics

    International Nuclear Information System (INIS)

    Dittmer, G.; Coyle, T.

    1998-01-01

    Some basic facts regarding derivatives in the electric power industry were discussed based on experiences in the natural gas industry. Derivatives were described as financial instruments that have no value of their own but derive their value from other assets, such as commodities. Futures and Options are the two forms of derivatives. Electricity as a commodity was characterized, and a historical parallel was drawn between deregulation in the natural gas and electric power industry. Short term and long term factors impacting the market and market dynamics impacting derivatives were identified. The latter include: (1) volatility, (2) liquidity, (3) correlations, and (4) price discovery. In contrast to the natural gas market, the electricity market is considered as lacking liquidity and in need of moving farther along the maturity timeline. The question of natural gas/ electricity convergence was also addressed refs., tabs., figs

  7. Gas transfer under breaking waves: experiments and an improved vorticity-based model

    Directory of Open Access Journals (Sweden)

    V. K. Tsoukala

    2008-07-01

    Full Text Available In the present paper a modified vorticity-based model for gas transfer under breaking waves in the absence of significant wind forcing is presented. A theoretically valid and practically applicable mathematical expression is suggested for the assessment of the oxygen transfer coefficient in the area of wave-breaking. The proposed model is based on the theory of surface renewal that expresses the oxygen transfer coefficient as a function of both the wave vorticity and the Reynolds wave number for breaking waves. Experimental data were collected in wave flumes of various scales: a small-scale experiments were carried out using both a sloping beach and a rubble-mound breakwater in the wave flume of the Laboratory of Harbor Works, NTUA, Greece; b large-scale experiments were carried out with a sloping beach in the wind-wave flume of Delft Hydraulics, the Netherlands, and with a three-layer rubble mound breakwater in the Schneideberg Wave Flume of the Franzius Institute, University of Hannover, Germany. The experimental data acquired from both the small- and large-scale experiments were in good agreement with the proposed model. Although the apparent transfer coefficients from the large-scale experiments were lower than those determined from the small-scale experiments, the actual oxygen transfer coefficients, as calculated using a discretized form of the transport equation, are in the same order of magnitude for both the small- and large-scale experiments. The validity of the proposed model is compared to experimental results from other researchers. Although the results are encouraging, additional research is needed, to incorporate the influence of bubble mediated gas exchange, before these results are used for an environmental friendly design of harbor works, or for projects involving waste disposal at sea.

  8. A validation study for the gas migration modelling of the compacted bentonite using existing experiment data

    International Nuclear Information System (INIS)

    Tawara, Y.; Mori, K.; Tada, K.; Shimura, T.; Sato, S.; Yamamoto, S.; Hayashi, H.

    2010-01-01

    Document available in extended abstract form only. After the field-scaled Gas Migration Test (GMT) was carried out at Grimsel Test Site (GTS) in Switzerland from 1997 through 2005, a study on advanced gas migration modelling has been conducted as a part of R and D programs of the RWMC (Radioactive Waste Management funding and Research Center) to evaluate long-term behaviour of the Engineered Barrier System (EBS) for the TRU waste disposal system in Japan. One of main objectives of this modelling study is to provide the qualified models and parameters in order to predict long-term gas migration behaviour in compacted bentonite. In addition, from a perspective of coupled THMC (Thermal, Hydrological, Mechanical and Chemical) processes, the specific processes which may have considerable impact to the gas migration behaviour are discussed by means of scoping calculations. Literature survey was conducted to collect experimental data related to gas migration in compacted bentonite in order to discuss an applicability of the existing gas migration models in the bentonite. The well-known flow rate controlled-gas injection experiment by Horseman, et al. and the pressure-controlled-gas injection test using several data with wide range of clay density and water content by Graham, et al, were selected. These literatures show the following characteristic behaviour of gas migration in high compacted and water-saturated bentonite. The observed gas flow rate from the outlet in the experiment by Horseman et al. was numerically reproduced by using the different conceptual models and computer codes, and then an applicability of the models and the identified key parameters such as relative permeability and capillary pressure were discussed. Helium gas was repeatedly injected into fully water-saturated and isotropically consolidated MX-80 bentonite (dry density: 1.6 Mg/m 3 ) in the experiment. One of the most important conclusions from this experiment is that it's impossible for

  9. Quantifying Reaeration Rates in Alpine Streams Using Deliberate Gas Tracer Experiments

    Directory of Open Access Journals (Sweden)

    Andrew Benson

    2014-04-01

    Full Text Available Gas exchange across the air-water interface is a critical process that maintains adequate dissolved oxygen (DO in the water column to support life. Oxygen reaeration rates can be accurately measured using deliberate gas tracers, like sulfur hexafluoride (SF6 or xenon (Xe. Two continuous release experiments were conducted in different creeks in the Sierra Nevada of California: Sagehen Creek in September, 2009, using SF6 and Martis Creek in August, 2012, using both SF6 and Xe. Measuring gas loss along the creek, which was approximated with the one-dimensional advection-dispersion equation, allows for the estimation of the SF6 or Xe reaeration coefficient (KSF6, KXe, which is converted to DO reaeration (KDO or K2 using Schmidt numbers. Mean KSF6 for upper and lower Sagehen and Martis Creeks were, respectively, 34 day−1, 37 day−1 and 33 day−1, with corresponding KDOs of 61 day−1, 66 day−1 and 47 day−1. In Martis Creek, KXe was slightly higher (21% than KSF6, but the calculated KDO from SF6 agreed with the calculated KDO from Xe within about 15%; this difference may be due to bubble-enhanced gas transfer. Established empirical equations of KDO using stream characteristics did a poor job predicting KDO for both creeks.

  10. Experience of iodine, caesium and noble gas release from AGR failures

    International Nuclear Information System (INIS)

    Chapman, C.J.; Harris, A.M.; Phillips, M.E.

    1985-01-01

    In the event of a fuel failure in an Advanced Gas Cooled Reactor (AGR), the quantity of fission products available for release to the environment is determined by the transport of fission products in the UO 2 fuel, by the possible retention of fission products in the fuel can interspace and by the deposition of fission products on gas circuit surfaces ('plate-out'). The fission products of principal radiological concern are radioactive caesium (Cs-137 and Cs-134) and iodine (principally I-131). Results are summarised of a number of experiments which were designed to study the release of these fission products from individual fuel failures in the prototype AGR at Windscale. Results are also presented of fission product release from failures in commercial AGRs. Comparisons of measured releases of caesium and iodine relative to the release of the noble gas fission products show that, for some fuel failures, there is a significant retention of caesium and iodine within the fuel can interspace. Under normal conditions circuit deposition reduces caesium and iodine gas concentrations by several orders of magnitude. Differing release behaviour of caesium and iodine from the failures is examined together with subsequent deposition within the sampling equipment. These observations are important factors which must be considered in developing an understanding of the mechanisms involved in circuit deposition. (author)

  11. Study of microstrip gas chambers for CMS experiment and measurement of the W boson mass in the DELPHI experiment

    International Nuclear Information System (INIS)

    Ripp-Baudot, I.

    2004-06-01

    In this document the author describes 3 fields of his research activities: first, the development and validation tests of micro-strip gas chambers for the CMS experiment; secondly, the measurements of the W boson mass and width by analysing the events: e + e - → W + W - → qq-bar qq-bar whose data have been collected in the DELPHI experiment (at the LEP-2 accelerator); and thirdly, the tagging of b-jets that is an essential tool for the study of the top quark. The last chapter is dedicated to what is expected from LHC experiments concerning the properties of the quark top: mass, spin, production and decay channels

  12. Neutral Transport Simulations of Gas Puff Imaging Experiments on Alcator C-Mod

    International Nuclear Information System (INIS)

    Stotler, D.P.; LaBombard, B.; Terry, J.L.; Zweben, S.J.

    2002-01-01

    Visible imaging of gas puffs has been used on the Alcator C-Mod tokamak to characterize edge plasma turbulence, yielding data that can be compared with plasma turbulence codes. Simulations of these experiments with the DEGAS 2 Monte Carlo neutral transport code have been carried out to explore the relationship between the plasma fluctuations and the observed light emission. By imposing two-dimensional modulations on the measured time-average plasma density and temperature profiles, we demonstrate that the spatial structure of the emission cloud reflects that of the underlying turbulence. However, the photon emission rate depends on the plasma density and temperature in a complicated way, and no simple scheme for inferring the plasma parameters directly from the light emission patterns is apparent. The simulations indicate that excited atoms generated by molecular dissociation are a significant source of photons, further complicating interpretation of the gas puff imaging results.Visibl e imaging of gas puffs has been used on the Alcator C-Mod tokamak to characterize edge plasma turbulence, yielding data that can be compared with plasma turbulence codes. Simulations of these experiments with the DEGAS 2 Monte Carlo neutral transport code have been carried out to explore the relationship between the plasma fluctuations and the observed light emission. By imposing two-dimensional modulations on the measured time-average plasma density and temperature profiles, we demonstrate that the spatial structure of the emission cloud reflects that of the underlying turbulence. However, the photon emission rate depends on the plasma density and temperature in a complicated way, and no simple scheme for inferring the plasma parameters directly from the light emission patterns is apparent. The simulations indicate that excited atoms generated by molecular dissociation are a significant source of photons, further complicating interpretation of the gas puff imaging results

  13. Integral analysis of debris material and heat transport in reactor vessel lower plenum

    International Nuclear Information System (INIS)

    Suh, K.Y.; Henry, R.E.

    1994-01-01

    An integral, fast-running, two-region model has been developed to characterize the debris material and heat transport in the reactor lower plenum under severe accident conditions. The debris bed is segregated into the oxidic pool and an overlying metallic layer. Debris crusts can develop on three surfaces: the top of the molten pool, the RPV wall, and the internal structures. To account for the decay heat generation, the crust temperature profile is assumed to be parabolic. The oxidic debris pool is homogeneously mixed and has the same material composition, and hence the same thermophysical properties, as the crusts, while the metallic constituents are assumed to rise to the top of the debris pool. Steady-state relationships are used to describe the heat transfer rates, with the assessment of solid or liquid state, and the liquid superheat in the pool being based on the average debris temperature. Natural convection heat transfer from the molten debris pool to the upper, lower and embedded crusts is calculated based on the pool Rayleigh number with the conduction heat transfer from the crusts being determined by the crust temperature profile. The downward heat flux is transferred to the lowest part of the RPV lower head through a crust-to-RPV contact resistance. The sideward heat flux is transferred to the upper regions of the RPV lower head as well as to the internal structures. The upward heat flux goes to the metal layer, water, or available heat sink structures above. Quenching due to water ingression is modeled separately from the energy transfer through the crust. The RPV wall temperature distribution and the primary system pressure are utilized to estimate challenges to the RPV integrity. ((orig.))

  14. Irradiation performance of U-Pu-Zr metal fuels for liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Tsai, H.; Cohen, A.B.; Billone, M.C.; Neimark, L.A.

    1994-10-01

    This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result's from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of ∼75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of ∼1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of ∼11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction

  15. Comparison between gas puffing and supersonic molecular beam injection in plasma density feedback experiments in EAST

    International Nuclear Information System (INIS)

    Zheng, Xingwei; Li, Jiangang; Hu, Jiansheng; Li, Jiahong; Ding, Rui; Cao, Bin; Wu, Jinhua

    2013-01-01

    To achieve desirable plasma density control, a supersonic molecular beam injection (SMBI) feedback control system has been developed recently for the EAST tokamak. The performance of the SMBI and gas puffing (GP) feedback systems were used and compared. The performance of pulse width mode is better than that of pulse amplitude mode when GP was used for density feedback control. During one-day experiments, the variation of gas input and wall retention can be clarified into two stages. In the first stage the retention ratio is as high as 80–90%, and the gas input is about an order of 10 22 D 2 . However, in the second stage, the retention ratio is at a range of 50–70%. The gas input of a single discharge is small and the net wall retention grows slowly. The results of the SMBI feedback control experiment was analyzed. The shorter delay time of SMBI makes it faster at feeding back control the plasma density. The result showed that, compared with GP, the gas input of SMBI was decreased ∼30% and the wall retention was reduced ∼40%. This shows SMBI's advantage for the long pulse high density discharges in EAST. (paper)

  16. Nuclear fuel element

    International Nuclear Information System (INIS)

    Grossman, L.N.; Levin, H.A.

    1975-01-01

    A nuclear fuel element has disposed therein an alloy having the essential components of nickel, titanium and zirconium, and the alloy reacts with water, water vapor and reactive gases at reactor ambient temperatures. The alloy is disposed in the plenum of the fuel element in the form of particles in a hollow gas permeable container having a multiplicity of openings of size smallr than the size of the particles. The container is preferably held in the spring in the plenum of the fuel element. (E.C.B.)

  17. Fire and Gas Detection in the LHC Experiments The Sniffer Project

    CERN Document Server

    Nunes, R W

    2001-01-01

    The LHC experiments, due to their complexity and size, present many safety challenges. Cryogenic gases are used in large quantities as well as certain flammable mixtures. The electrical power involved calls for analysis of the fire risks. Access is restricted to the minimum and environmental conditions are extremely harsh, due to strong magnetic fields and ionising radiation. This paper will describe the Combined Fire/Gas/Oxygen deficiency Detection systems proposed for inside the ATLAS and CMS Experiments and possibly for the two others, if they deem it necessary. The requirements of the experiments and the development and implementation of such a system will be discussed. In parallel, commercial procedures to implement these systems by industry shall be described, taking into consideration that a previous development has already been undertaken by CERN for the LEP experiments. The stage is set for inter-divisional collaboration in a project of utmost importance for the safety of people and protection of the...

  18. LOFT advanced fuel rod instrumentation development

    International Nuclear Information System (INIS)

    Billeter, T.R.; Brown, R.L.; Chan, A.I.Y.; Day, C.K.; Meyers, S.C.; Sheen, E.M.; Stringer, J.L.

    1978-01-01

    Advanced fuel rod instrumentation for the Loss of Fluid Test (LOFT) reactor is being developed by the Hanford Engineering Development Laboratory for the Nuclear Regulatory Commission. This effort calls for development of sensors to measure fuel rod axial motion, fuel centerline temperature (to 2200 0 C), fuel rod plenum gas pressure (to 2500 psig), and plenum gas temperature (to 1500 0 F). A parallel test and evaluation of several modified commercial sensors was undertaken and will result in commercial availability of the final qualified sensors. Necessary test facilities were prepared for the development and evaluation effort. Tests to date indicate a three coil Linear Variable Differential Transformer (LVDT), operated from temperature compensating signal source and processing electronics, will meet the desired requirements

  19. Reactor container

    International Nuclear Information System (INIS)

    Fukazawa, Masanori.

    1991-01-01

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

  20. A non-destructive, ultrasonic method for the determination of internal pressure and gas composition in an LWR fuel rod on-going and future programme

    International Nuclear Information System (INIS)

    Ferrandis, J.; Leveque, G.; Villard, J.

    2006-01-01

    Several possible non-destructive methods have been investigated in the past to measure the internal gas pressure e.g., measurement of 85 Kr directly, or after accumulation in the plenum by freezing with liquid nitrogen. However no satisfactory resolution to the problem has been found, so at present there is no rapid and accurate method of determining the fission gas pressure in a fuel rod without puncturing the cladding. This procedure is time-consuming and expensive and as a consequence a relatively small number of measurements are generally made compared with the number of fuel rods irradiated. In this paper it is proposed a new method for the measurement of pressure that is: Non-destructive; Non-invasive (i.e., allows re-irradiation of the measured rod); Easy to operate - directly in the reactor pool; Can be used on the critical path; Is inexpensive compared with the methods currently in use. This method is also being adapted to the on line measurement of fission gas release on fuel irradiation in research reactors. This method is based on the application of acoustic technology

  1. Instellar Gas Experiment (IGE): Testing interstellar gas particles to provide information on the processes of nucleosynthesis in the big bang stars and supernova

    Science.gov (United States)

    Lind, Don

    1985-01-01

    The Interstellar Gas Experiment (IGE) is designed to collect particles of the interstellar gas - a wind of interstellar media particles moving in the vicinity of the solar system. These particles will be returned to earth where the isotopic ratios of the noble gases among these particles will be measured. IGE was designed and programmed to expose 7 sets of six copper-beryllium metallic collecting foils to the flux of neutral interstellar gas particles which penetrate the heliosphere to the vicinity of the earth's orbit. These particles are trapped in the collecting foils and will be returned to earth for mass-spectrographic analysis when Long Duration Exposure Facility (LDEF) on which IGE was launched, is recovered.

  2. High density experiments with gas puffing and ECRH in T-10

    International Nuclear Information System (INIS)

    Esipchuk, Yu V; Kirneva, N A; Borschegovskij, A A; Chistyakov, V V; Denisov, V Ph; Dremin, M M; Gorbunov, E P; Grashin, S A; Kalupin, D V; Khimchenko, L N; Khramenkov, A V; Kirnev, G S; Krilov, S V; Krupin, V A; Myalton, T B; Pavlov, Yu D; Piterskij, V V; Ploskirev, G N; Poznyak, V I; Roy, I N; Shelukhin, D A; Skosyrev, Yu V; Trukhin, V M; Trukhina, E V; Vershkov, V A; Veschev, E A; Volkov, V V; Zhuravlev, V A

    2003-01-01

    High density experiments were carried out in T-10 with gas puffing and electron cyclotron resonance heating (with absorbed power value up to 1.4 MW) with oblique and perpendicular power launch. Densities exceeding the Greenwald limit (n Gw ) by up to a factor of 1.8 were achieved in a regime with a high value of the edge safety factor at the current flat-top, q(a)≅8.2. The decrease of q(a) to a value of 3 led to the reduction of the ratio ( n-bar e ) lim /n Gw to 1. Confinement degradation with density increase was not significant up to the density limit. However, the typical T-10 linear increase of energy confinement time with density saturates at n-bar e ≥0.6n Gw . This saturation is the result of the development of an additional transport in the electron heat channel. However, the saturated τ E values exceeded the ITER L-mode scaling predictions by up to a factor of 1.2 and were close to the value predicted by the ITER H-mode scaling. Effect of the strong gas puffing on the plasma confinement and experiments with neon seeding are also discussed in this paper

  3. Experiments on vertical gas-liquid pipe flows using ultrafast X-ray tomography

    Energy Technology Data Exchange (ETDEWEB)

    Banowski, M.; Beyer, M.; Lucas, D.; Hoppe, D.; Barthel, F. [Helmholtz-Zentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung

    2016-12-15

    For the qualification and validation of two-phase CFD-models for medium and large-scale industrial applications dedicated experiments providing data with high temporal and spatial resolution are required. Fluid dynamic parameter like gas volume fraction, bubble size distribution, velocity or turbulent kinetic energy should be measured locally. Considering the fact, that the used measurement techniques should not affect the flow characteristics, radiation based tomographic methods are the favourite candidate for such measurements. Here the recently developed ultrafast X-ray tomography, is applied to measure the local and temporal gas volume fraction distribution in a vertical pipe. To obtain the required frame rate a rotating X-ray source by a massless electron beam and a static detector ring are used. Experiments on a vertical pipe are well suited for development and validation of closure models for two-phase flows. While vertical pipe flows are axially symmetrically, the boundary conditions are well defined. The evolution of the flow along the pipe can be investigated as well. This report documents the experiments done for co-current upwards and downwards air-water and steam-water flows as well as for counter-current air-water flows. The details of the setup, measuring technique and data evaluation are given. The report also includes a discussion on selected results obtained and on uncertainties.

  4. Numerical analysis of experiments with gas injection into liquid metal coolant

    International Nuclear Information System (INIS)

    Usov, E V; Lobanov, P D; Pribaturin, N A; Mosunova, N A; Chuhno, V I; Kutlimetov, A E

    2016-01-01

    Presented paper contains results of a numerical analysis of experiments with gas injection in water and liquid metal which have been performed at the Institute of Thermophysics Russian Academy of Science (IT RAS). Obtained experimental data are very important to predict processes that take place in the BREST-type reactor during the hypothetical accident with damage of the steam generator tubes, and may be used as a benchmark to validate thermo-hydraulic codes. Detailed description of models to simulate transport of gas phase in a vertical liquid column is presented in a current paper. Two-fluid model with closing relation for wall friction and interface friction coefficients was used to simulate processes which take place in a liquid during injection of gaseous phase. It has being shown that proposed models allow obtaining a good agreement between experimental data and calculation results. (paper)

  5. Characterization of a plasma produced using a high power laser with a gas puff target for x-ray laser experiments

    International Nuclear Information System (INIS)

    Fiedorowicz, H.; Bartnik, A.; Gac, K.; Parys, P.; Szczurek, M.; Tyl, J.

    1995-01-01

    A high temperature, high density plasma can be produced by using a nanosecond, high-power laser with a gas puff target. The gas puff target is formed by puffing a small amount of gas from a high-pressure reservoir through a nozzle into a vacuum chamber. In this paper we present the gas puff target specially designed for x-ray laser experiments. The solenoid valve with the nozzle in the form of a slit 0.3-mm wide and up to 40-mm long, allows to form an elongated gas puff suitable for the creation of an x-ray laser active medium by its perpendicular irradiation with the use of a laser beam focused to a line. Preliminary results of the experiments on the laser irradiation of the gas puff targets, produced by the new valve, show that hot plasma suitable for x-ray lasers is created

  6. Molybdenum-UO2 cerment irradiation at 1145 K

    Science.gov (United States)

    Mcdonald, G.

    1971-01-01

    Two molybdenum-UO2 cermet fuel pins were fission heated in a helium-cooled loop at a temperature of 1145 K and to a total burnup of 5.3 % of the U-235. After irradiation the fuel pins were measured to check dimensional stability, punctured at the plenums to determine fission gas release, and examined metallographically to determine the effect of irradiation. Burnup was determined in several sections of the fuel pin. The results of the postirradiation examination indicated: (1) There was no visible change in the fuel pins on irradiation under the above conditions. (2) The maximum swelling of the fuel pins was less than 1%. (3) There was no migration of UO2 and no visible interaction between the molybdenum and the UO2. (4) Approximately 12% of the fission gas formed was released from the cermet cone into the gas plenum.

  7. The Cloud Ice Mountain Experiment (CIME) 1998: experiment overview and modelling of the microphysical processes during the seeding by isentropic gas expansion

    Science.gov (United States)

    Wobrock, Wolfram; Flossmann, Andrea I.; Monier, Marie; Pichon, Jean-Marc; Cortez, Laurent; Fournol, Jean-François; Schwarzenböck, Alfons; Mertes, Stephan; Heintzenberg, Jost; Laj, Paolo; Orsi, Giordano; Ricci, Loretta; Fuzzi, Sandro; Brink, Harry Ten; Jongejan, Piet; Otjes, René

    The second field campaign of the Cloud Ice Mountain Experiment (CIME) project took place in February 1998 on the mountain Puy de Dôme in the centre of France. The content of residual aerosol particles, of H 2O 2 and NH 3 in cloud droplets was evaluated by evaporating the drops larger than 5 μm in a Counterflow Virtual Impactor (CVI) and by measuring the residual particle concentration and the released gas content. The same trace species were studied behind a round jet impactor for the complementary interstitial aerosol particles smaller than 5 μm diameter. In a second step of experiments, the ambient supercooled cloud was converted to a mixed phase cloud by seeding the cloud with ice particles by the gas release from pressurised gas bottles. A comparison between the physical and chemical characteristics of liquid drops and ice particles allows a study of the fate of the trace constituents during the presence of ice crystals in the cloud. In the present paper, an overview is given of the CIME 98 experiment and the instrumentation deployed. The meteorological situation during the experiment was analysed with the help of a cloud scale model. The microphysics processes and the behaviour of the scavenged aerosol particles before and during seeding are analysed with the detailed microphysical model ExMix. The simulation results agreed well with the observations and confirmed the assumption that the Bergeron-Findeisen process was dominating during seeding and was influencing the partitioning of aerosol particles between drops and ice crystals. The results of the CIME 98 experiment give an insight on microphysical changes, redistribution of aerosol particles and cloud chemistry during the Bergeron-Findeisen process when acting also in natural clouds.

  8. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    Energy Technology Data Exchange (ETDEWEB)

    Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

  9. Experiences of small-scale consumers in the market for natural gas. Visible results of liberalization

    International Nuclear Information System (INIS)

    Neeleman, J.

    2005-01-01

    The gas market, as the electricity market, has been fully liberalized for all consumers in the Netherlands since 1 July 2004. Was the market ready for this? What main changes have taken place? Health care institutions in the province of Zeeland have now had their first experiences with the liberalized gas market and the outcome has been a saving of 17%. A study of the tendering process for 86 care institutions reveals that not all the gas supply companies were ready to sign contracts with a combination of smaller and larger consuming organizations. Another conclusion is that these consumers expended some effort learning about the gas market, which appears to be nontransparent to outsiders. The latter is a striking observation because costs are the most important criterion for institutional consumers [nl

  10. Stratospheric Aerosol and Gas Experiment III on the International Space Station (SAGE III/ISS)

    Science.gov (United States)

    Gasbarre, Joseph; Walker, Richard; Cisewski, Michael; Zawodny, Joseph; Cheek, Dianne; Thornton, Brooke

    2015-01-01

    The Stratospheric Aerosol and Gas Experiment III on the International Space Station (SAGE III/ISS) mission will extend the SAGE data record from the ideal vantage point of the International Space Station (ISS). The ISS orbital inclination is ideal for SAGE measurements providing coverage between 70 deg north and 70 deg south latitude. The SAGE data record includes an extensively validated data set including aerosol optical depth data dating to the Stratospheric Aerosol Measurement (SAM) experiments in 1975 and 1978 and stratospheric ozone profile data dating to the Stratospheric Aerosol and Gas Experiment (SAGE) in 1979. These and subsequent data records, notably from the SAGE II experiment launched on the Earth Radiation Budget Satellite in 1984 and the SAGE III experiment launched on the Russian Meteor-3M satellite in 2001, have supported a robust, long-term assessment of key atmospheric constituents. These scientific measurements provide the basis for the analysis of five of the nine critical constituents (aerosols, ozone (O3), nitrogen dioxide (NO2), water vapor (H2O), and air density using O2) identified in the U.S. National Plan for Stratospheric Monitoring. SAGE III on ISS was originally scheduled to fly on the ISS in the same timeframe as the Meteor-3M mission, but was postponed due to delays in ISS construction. The project was re-established in 2009.

  11. Intense ion beam transport in magnetic quadrupoles: Experiments on electron and gas effects

    International Nuclear Information System (INIS)

    Seidl, P.A.; Molvik, A.W.; Bieniosek, F.M.; Cohen, R.H.; Faltens, A.; Friedman, A.; Kireef Covo, M.; Lund, S.M.; Prost, L.; Vay, J-L.

    2004-01-01

    Heavy-ion induction linacs for inertial fusion energy and high-energy density physics have an economic incentive to minimize the clearance between the beam edge and the aperture wall. This increases the risk from electron clouds and gas desorbed from walls. We have measured electron and gas emission from 1 MeV K + incident on surfaces near grazing incidence on the High-Current Experiment (HCX) at LBNL. Electron emission coefficients reach values >100, whereas gas desorption coefficients are near 10 4 . Mitigation techniques are being studied: A bead-blasted rough surface reduces electron emission by a factor of 10 and gas desorption by a factor of 2. We also discuss the results of beam transport (of 0.03-0.18 A K + ) through four pulsed room-temperature magnetic quadrupoles in the HCX at LBNL. Diagnostics are installed on HCX, between and within quadrupole magnets, to measure the beam halo loss, net charge and expelled ions, from which we infer gas density, electron trapping, and the effects of mitigation techniques. A coordinated theory and computational effort has made significant progress towards a self-consistent model of positive-ion beam and electron dynamics. We are beginning to compare experimental and theoretical results

  12. Application of Gas Chromatographic analysis to RPC detectors in the ATLAS experiment at CERN-LHC

    CERN Document Server

    De Asmundis, R

    2007-01-01

    Starting from 2007 a large number (1200) Resistive Plate Chambers (RPC) detectors will be used as muon trigger detectors in the ATLAS Experiment at CERN-LHC accelerator. RPC are gaseous detector in which the quality and the stability of the gas mixture as well as the design of the gas supplying system, play a fundamental role in their functioning. RPC are foreseen to work more than ten years in the high radiation environment of ATLAS and the gas mixture acts really as a "lifeguard" for the detectors. For this reason a great attention has been devoted to the gas studies in order to optimize RPC performance, robustness and reliability in a high radiation environment. In this paper we describe the work done to decide how to supply and control in an optimal way the gas to the detectors, in order to ensure their best performance for a long time. The activity, based on Gas Chromatographic (GC) analysis, has been carried on a sample of final RPC working in radiation conditions much more intense than those foreseen f...

  13. Open access to the natural gas transport system. Experiences in North America and developing trends in Europe

    Energy Technology Data Exchange (ETDEWEB)

    Bundgaard-Joergensen, U; Hopper, R J

    1988-09-01

    The treatise describes potential market forces which could evoke changes in the structure of the European gas industry or in its types of contract. It shows that a price differentiation at the borehole may lead to an increase in natural gas deliveries for the European markets. A study of the development of the North American gas industry over the last few decades supports this expectation. The treatise ends with the statement that an application of the North American experiences to the European gas industry is unlikely, but it does not rule out the possibility of market forces or the EEC Commission creating a basis for similar structural reforms in the European gas industry.

  14. Studies of Flow in Ionized Gas: Historical Perspective, Contemporary Experiments, and Applications

    International Nuclear Information System (INIS)

    Popovic, S.; Vuskovic, L.

    2007-01-01

    Since the first observations that a very small ionized fraction (order of 1 ppm) could strongly affect the gas flow, numerous experiments with partially or fully wall-free discharges have demonstrated the dispersion of shock waves, the enhancement of lateral forces in the flow, the prospects of levitation, and other aerodynamic effects with vast potential of application. A review of physical effects and observations are given along with current status of their interpretation. Special attention will be given to the physical problems of energy efficiency in generating wall-free discharges and the phenomenology of filamentary discharges. Comments and case examples are given on the current status of availability of necessary data for modelling and simulation of the aerodynamic phenomena in weakly ionized gas

  15. Analysis of the VVER Standard Problem INSC-PSBV1 '11% Coolant Leak from Upper Plenum' with RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    Melikhov, O.; Melikhov, V.; Parfenov, Yu.; Gavritenkova, O.; Lipatov, I.; Elkin, I.; Bayless, P.

    2004-01-01

    Analyses of a loss-of-coolant experiment carried out at the PSB-VVER test facility with the RELAP5/MOD3.2 code have been performed independently by analysts at the Electrogorsk Research and Engineering Center (EREC) and the Idaho National Engineering and Environmental Laboratory (INEEL). The PSB-VVER facility is a full-height scale model of a VVER 1000 reactor that is approximately 1/300 scale in volume and power. VVER Standard Problem INSC-PSBV1 represents an 11% leak from the upper plenum of the PSB-VVER facility, simulating the rupture of one of the accumulator injection lines. The safety-significant thermalhydraulic phenomena occurring in VVER type reactors addressed by this experiment were identified in the test validation matrix. Most of the phenomena of the validation matrix were reasonably simulated by RELAP5/MOD3.2 in both calculations. The major differences between the test and the calculations were the timing of the core heatup, and the thermal response to the accumulator injection cycles in both calculations. The INEEL calculation had a more extensive axial heatup, with most of the core experiencing small heat-ups. The accumulator injection was more effective in quenching the core in the test than in the INEEL calculation. This difference is attributed to the liquid distribution in the core, rather than to the heat transfer models in the code. The code calculation had a more uniform axial distribution of the liquid in the core, and the accumulator injection did not have much impact on the core liquid inventory. In the EREC calculation, only one heatup of the cladding temperature was observed for upper and middle section of the fuel rods before the final heatup. The small heat-ups were not reproduced in EREC calculation. The difference could be attributed to differences in liquid distribution, namely the core region in the EREC calculation contains more liquid over most of the transient than in the experiment. The distribution of liquid in the core in

  16. Validation of the new filters configuration for the RPC gas systems at LHC experiments

    CERN Document Server

    Mandelli, Beatrice; Guida, Roberto; Hahn, Ferdinand; Haider, Stefan

    2012-01-01

    Resistive Plate Chambers (RPCs) are widely employed as muon trigger systems at the Large Hadron Collider (LHC) experiments. Their large detector volume and the use of a relatively expensive gas mixture make a closed-loop gas circulation unavoidable. The return gas of RPCs operated in conditions similar to the experimental background foreseen at LHC contains large amount of impurities potentially dangerous for long-term operation. Several gas-cleaning agents, characterized during the past years, are currently in use. New test allowed understanding of the properties and performance of a large number of purifiers. On that basis, an optimal combination of different filters consisting of Molecular Sieve (MS) 5Å and 4Å, and a Cu catalyst R11 has been chosen and validated irradiating a set of RPCs at the CERN Gamma Irradiation Facility (GIF) for several years. A very important feature of this new configuration is the increase of the cycle duration for each purifier, which results in better system stabilit...

  17. Gas puff modulation experiments in Tore Supra

    International Nuclear Information System (INIS)

    Haas, J.C.M. de; Devynck, P.; Dudok de Wit, T.; Garbet, X.; Gil, C.; Harris, G.; Laviron, C.; Martin, G.

    1993-01-01

    Experiments with a modulation of the gas puff have been done in Tore Supra with the aim to investigate the transport of particles and heat. The target plasma is ohmically heated, sawtoothing with frequencies between 12 and 20 Hz, deuterium for both the plasma and the injection, and with various densities, rising in a series of shots. Both the diffusion coefficient and the pinch velocity for the particle transport were determined using an harmonic modulation. The method gives reasonable results, even for small perturbations, and the obtained values are able to reproduce the stationary values. The heat flow carried by electrons also shows a modulation. The part of the modulation which is not caused by the density can in principle be used to discriminate diffusive and convective terms in the heat flux. An ion temperature profile calculated with empirically determined value of heat diffusivity reproduces the slow evolution of the total kinetic energy. 6 figs., 7 refs

  18. Computer simulation of void formation in residual gas atom free metals by dual beam irradiation experiments

    International Nuclear Information System (INIS)

    Shimomura, Y.; Nishiguchi, R.; La Rubia, T.D. de; Guinan, M.W.

    1992-01-01

    In our recent experiments (1), we found that voids nucleate at vacancy clusters which trap gas atoms such as hydrogen and helium in ion- and neutron-irradiated copper. A molecular dynamics computer simulation, which implements an empirical embedded atom method to calculate forces that act on atoms in metals, suggests that a void nucleation occurs in pure copper at six and seven vacancy clusters. The structure of six and seven vacancy clusters in copper fluctuates between a stacking fault tetrahedron and a void. When a hydrogen is trapped at voids of six and seven vacancy, a void can keep their structure for appreciably long time; that is, the void do not relax to a stacking fault tetrahedron and grows to a large void. In order to explore the detailed atomics of void formation, it is emphasized that dual-beam irradiation experiments that utilize beams of gas atoms and self-ions should be carried out with residual gas atom free metal specimens. (author)

  19. Gas-filled Rugby hohlraum energetics and implosions experiments on OMEGA

    Science.gov (United States)

    Casner, Alexis; Philippe, F.; Tassin, V.; Seytor, P.; Monteil, M. C.; Villette, B.; Reverdin, C.

    2010-11-01

    Recent experiments [1,2] have validated the x-ray drive enhancement provided by rugby-shaped hohlraums over cylinders in the indirect drive (ID) approach to inertial confinement fusion (ICF). This class of hohlraum is the baseline design for the Laser Mégajoule program, is also applicable to the National Ignition Facility and could therefore benefit ID Inertial Fusion Energy studies. We have carried out a serie of energetics and implosions experiments with OMEGA ``scale 1'' rugby hohlraums [1,2]. For empty hohlraums these experiments provide complementary measurements of backscattered light along 42 cone, as well as detailed drive history. In the case of gas-filled rugby hohlraums we have also study implosion performance (symmetry, yield, bangtime, hotspot spectra...) using a high contrast shaped pulse leading to a different implosion regime and for a range of capsule convergence ratios. These results will be compared with FCI2 hydrocodes calculations and future experimental campaigns will be suggested. [4pt] [1] F. Philippe et al., Phys. Rev. Lett. 104, 035004 (2010). [0pt] [2] H. Robey et al., Phys. Plasnas 17, 056313 (2010).

  20. Fission gas release modelling: developments arising from instrumented fuel assemblies, out-of-pile experiments and microstructural observations

    International Nuclear Information System (INIS)

    Leech, N.A.; Smith, M.R.; Pearce, J.H.; Ellis, W.E.; Beatham, N.

    1990-01-01

    This paper reviews the development of fission gas release modelling in thermal reactor fuel (both steady-state and transient) and in particular, illustrates the way in which experimental data have been, and continue to be, the main driving force behind model development. To illustrate this point various aspects of fuel performance are considered: temperature calculation, steady-state and transient fission gas release, grain boundary gas atom capacity and microstructural phenomena. The sources of experimental data discussed include end-of-life fission gas release measurements, instrumented fuel assemblies (e.g. rods with internal pressure transducers, fuel centre thermocouples), swept capsule experiments, out-of-pile annealing experiments and microstructural techniques applied during post-irradiation evaluation. In the case of the latter, the benefit of applying many observation and analysis techniques on the same fuel samples (the approach adopted at NRL Windscale) is emphasized. This illustrates a shift of emphasis in the modelling field from the development of large, complex thermo-mechanical computer codes to the assessment of key experimental data in order to develop and evaluate sub-models which correctly predict the observed behaviour. (author)

  1. Rheinbraun`s experience in hot gas cleaning

    Energy Technology Data Exchange (ETDEWEB)

    Renzenbrink, W.; Wischnewski, R. [Rheinbraun AG, Koeln (Germany)

    1998-11-01

    For the introduction of modern types of power stations like IGCC, PCFBC, etc. the application of a functional hot gas filter is of essential importance. A hot gas filter with two tiers for dry and complete dedusting of the entire raw gas flow of 53,000 m{sup 3}(STP)/h was started up in 1993 in the High Temperature Winkler (HTW) coal gasification demonstration plant in Hurth/Berrenrath near Cologne, Germany. The operational data of the filter are a pressure of 10 bar and a temperature of 270{degree}C. The filter was supplied by the `LLB` company and is characterised by the principle of upright arrangement of the ceramic filter elements. During nearly 8,000 h of plant operation up to September 1995 the filter showed stable and safe operation, a separation efficiency of {gt}99.98%, a 21% reduction in filtration surface, reduction in cleaning gas requirement by factor 10, reduction in cleaning gas pressure to 16 bar and a significant reduction in maintenance and operating costs. The resultant clean gas dust content was {lt} 3 mg/m{sup 3}(STP) compared to the design value of 5 mg/m{sup 3}(STP). In a test to the limit of operation one failure occurred when 20 candles broke. In order to yield larger filtering surfaces in very large filter units, e.g. for IGCCs, without using more than one filter the multistage design is the only sensible solution. Prior to industrial-scale application such a system has to be tested. Therefore the two-tier filter was converted into a three-tier type with separate filter modules at the end of 1995. After another 5,400 h of plant operation this three-tier filter shows safe and stable operation with a clean gas dust content of {lt} 2 mg/m{sup 3}(STP). 3 refs., 5 figs., 1 tab.

  2. Operation experiences of landfill gas engines; Motorer foer deponigas - Tillgaenglighet och drifterfarenheter

    Energy Technology Data Exchange (ETDEWEB)

    Dejfors, Charlotte; Grimberger, Goeran [AaF-Energikonsult Stockholm AB (Sweden)

    2000-06-01

    The gas that is obtained from landfilled waste is produced by bacteria that digest organic material in an anaerobic environment. Landfill gas consists mainly of methane, carbon dioxide and water vapour. It may be used either as auxiliary fuel in boilers close to the landfill or to generate electricity by means of a gas engine. Several plants where landfill gas is used in gas engines have had serious problems, a. o. with burned exhaust valves. These problems may occur already after a short period of operation, which influences the profitability. The purposes of the project reported were to collect operational experience in Sweden with engines using landfill gas as fuel, to identify which problems there are and which actions or improvements have been implemented in order to correct for these problems. Today, there are 9 facilities where landfill gas is used to fuel a total of 13 gas engines. In addition, there is an engine in Goeteborg which has scarcely been in operation after its installation because there is not enough gas. Contact has been taken with all these facilities. Many have pointed out that the gas engines are sensitive in the vicinity of maximum load, where the control system requires an even gas flow and a stable composition of the gas. A counter-measure in the facilities is to avoid running the engine at full load. All engines are equipped with a lean-NO{sub x} system in order to minimise NO{sub x} emissions. Many have remarked that the lean-NO{sub x} system shuts the engine off when emissions exceed the allowed limits. There is a consensus that spark plugs and ignition cables have created operational problems. These have been changed more frequently than originally expected. Another problem, which has caused operational problems and a need for maintenance, is deposits mainly in the combustion chamber, in valves and cylinder heads. Deposits and high exhaust gas temperature have led to burnt exhaust gas valves and cylinder heads on half of the engines

  3. Precalculation of the fission gas behaviour in the MOL 7C/6 experiment with the LAKU model

    International Nuclear Information System (INIS)

    Vaeth, L.

    1988-03-01

    The fission gas behaviour in the planned experiment MOL 7C/6 is simulated with the Karlsruhe model LAKU, employing temperatures calculated with the pin behaviour model TRANSURANUS. Two different modes of experimental flow blockage simulation are investigated and compared to an estimated fission gas behaviour during a realistic blockage build-up. The results indicate, that the start-up procedure leading to greatly reduced fission gas content is the more realistic one. Details of the calculations and their results are presented in the report

  4. Simulation and uncertainties of the heat transfer from a heat-generating DEBRIS bed in the lower plenum

    International Nuclear Information System (INIS)

    Schaaf, K.; Trambauer, K.

    1999-01-01

    The findings of the TMI-2 post-accident analyses indicated that internal cooling mechanisms may have a considerable potential to sustain the vessel integrity after a relocation of core material to the lower plenum, provided that water is continuously available in the RPV. Numerous analytical and experimental research activities are currently underway in this respect. This paper illustrates some major findings of the experimental work on internal cooling mechanisms and describes the limitations and the uncertainties in the simulation of the heat transfer processes. Reference is made especially to the joint German DEBRIS/ RPV research program, which encompasses the experimental investigation of the thermal-hydraulics in gaps, of the heat transfer within a particulate debris bed, and of the high temperature performance of vessel steel, as well as the development of simulation models for the heat transfer in the lower head and the structural response of the RPV. In particular, the results of uncertainty and sensitivity analyses are presented, which have been carried out at GRS using an integral model that describes the major phenomena governing the long-term integrity of the reactor vessel. The investigation of a large-scale relocation indicated that the verification of a gap cooling mechanism as an inherent mechanism is questionable in terms of a stringent probabilistic uncertainty criterion, as long as the formation of a large molten pool cannot be excluded. (author)

  5. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Grover, S. Blaine

    2009-01-01

    The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy's lead laboratory for nuclear energy development. The ATR is one of the world's premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  6. BISON Theory Manual The Equations behind Nuclear Fuel Analysis

    International Nuclear Information System (INIS)

    Hales, J. D.; Williamson, R. L.; Novascone, S. R.; Pastore, G.; Spencer, B. W.; Stafford, D. S.; Gamble, K. A.; Perez, D. M.; Liu, W.

    2016-01-01

    BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact, and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.

  7. Flow instability tests for a particle bed reactor nuclear thermal rocket fuel element

    Science.gov (United States)

    Lawrence, Timothy J.

    1993-05-01

    Recent analyses have focused on the flow stability characteristics of a particle bed reactor (PBR). These laminar flow instabilities may exist in reactors with parallel paths and are caused by the heating of the gas at low Reynolds numbers. This phenomena can be described as follows: several parallel channels are connected at the plenum regions and are stabilized by some inlet temperature and pressure; a perturbation in one channel causes the temperature to rise and increases the gas viscosity and reduces the gas density; the pressure drop is fixed by the plenum regions, therefore, the mass flow rate in the channel would decrease; the decrease in flow reduces the ability to remove the energy added and the temperature increases; and finally, this process could continue until the fuel element fails. Several analyses based on different methods have derived similar curves to show that these instabilities may exist at low Reynolds numbers and high phi's ((Tfinal Tinitial)/Tinitial). These analyses need to be experimentally verified.

  8. BISON Theory Manual The Equations behind Nuclear Fuel Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Williamson, R. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Novascone, S. R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Pastore, G. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spencer, B. W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stafford, D. S. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Perez, D. M. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Liu, W. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    BISON is a finite element-based nuclear fuel performance code applicable to a variety of fuel forms including light water reactor fuel rods, TRISO particle fuel, and metallic rod and plate fuel. It solves the fully-coupled equations of thermomechanics and species diffusion, for either 2D axisymmetric or 3D geometries. Fuel models are included to describe temperature and burnup dependent thermal properties, fission product swelling, densification, thermal and irradiation creep, fracture, and fission gas production and release. Plasticity, irradiation growth, and thermal and irradiation creep models are implemented for clad materials. Models are also available to simulate gap heat transfer, mechanical contact, and the evolution of the gap/plenum pressure with plenum volume, gas temperature, and fission gas addition. BISON is based on the MOOSE framework and can therefore efficiently solve problems using standard workstations or very large high-performance computers. This document describes the theoretical and numerical foundations of BISON.

  9. Experiments in stratified gas-liquid pipe flow

    NARCIS (Netherlands)

    Birvalski, M.

    2015-01-01

    The growing demand for energy in the future will necessitate the production of natural gas from fields which are located farther offshore, in deep water and in very cold environments. This will confront us with difficulties in ensuring continuous production of the fluids (natural gas, condensate and

  10. Experimental study on in-vessel debris coolability during severe accident

    International Nuclear Information System (INIS)

    Kim, S. B.; Park, R. J.; Kim, H. D.

    2002-05-01

    A research program, called SONATA-IV(Simulation of Naturally Arrested Thermal Attack In-Vessel), has been performed to verify the gap cooling mechanism of corium in the lower plenum, and to develop management and mitigation strategies under severe accident conditions. For the proof-of-principles experiment, the LAVA(Lower-plenum Arrested Vessel Attack) experiments have been performed to gather proof of gap formation and to evaluate the gap effect on in-vessel cooling, using Al 2 O 3 /Fe (or Al 2 O 3 only) thermite melt as corium simulant. And also the CHFG(Critical Heat Flux in Gap) experiments have been performed to measure the critical power and to investigate the inherent cooling mechanism in the hemispherical narrow gap. In addition to the experiments, LILAC code was developed to analyze and predict the thermo-hydraulic phenomena of the corium relocated in the reactor lower plenum. It could be found from the LAVA and CHFG experimental results that continuous gap ranged from 1 to 5 mm was formed and that maximum heat removal capacity through a gap is a key factor in determining the potentials of the integrity of the vessel. After all the possibility of IVR(In-Vessel corium Retention) through gap cooling highly depends on the melt relocated into the lower plenum and the gap size. So, feasibility experiments have been performed for the assessment of improved IVR concepts using an internal engineered gap device and a dual strategy of In/Ex-vessel cooling using the LAVA facility. It is preliminarily concluded that these cooling measures lead to an enhanced cooling of the corium in the lower plenum of the reactor vessel. The additional studies will be performed to verify the quantitative heat removal capacity for these cooling measures in the 2nd phase of mid- and long term project period

  11. Developments for transactinide chemistry experiments behind the gas-filled separator TASCA

    Energy Technology Data Exchange (ETDEWEB)

    Even, Julia

    2011-12-13

    Topic of this thesis is the development of experiments behind the gas-filled separator TASCA (TransActinide Separator and Chemistry Apparatus) to study the chemical properties of the transactinide elements. In the first part of the thesis, the electrodepositions of short-lived isotopes of ruthenium and osmium on gold electrodes were studied as model experiments for hassium. From literature it is known that the deposition potential of single atoms differs significantly from the potential predicted by the Nernst equation. This shift of the potential depends on the adsorption enthalpy of therndeposited element on the electrode material. If the adsorption on the electrode-material is favoured over the adsorption on a surface made of the same element as the deposited atom, the electrode potential is shifted to higher potentials. This phenomenon is called underpotential deposition. Possibilities to automatize an electro chemistry experiment behind the gas-filled separator were explored for later studies with transactinide elements. The second part of this thesis is about the in-situ synthesis of transition-metal-carbonyl complexes with nuclear reaction products. Fission products of uranium-235 and californium-249 were produced at the TRIGA Mainz reactor and thermalized in a carbon-monoxide containing atmosphere. The formed volatile metal-carbonyl complexes could be transported in a gas-stream. Furthermore, short-lived isotopes of tungsten, rhenium, osmium, and iridium were synthesised at the linear accelerator UNILAC at GSI Helmholtzzentrum fuer Schwerionenforschung, Darmstadt. The recoiling fusion products were separated from the primary beam and the transfer products in the gas-filled separator TASCA. The fusion products were stopped in the focal plane of TASCA in a recoil transfer chamber. This chamber contained a carbon-monoxide - helium gas mixture. The formed metal-carbonyl complexes could be transported in a gas stream to various experimental setups. All

  12. Transient fission gas release during direct electrical heating experiments

    International Nuclear Information System (INIS)

    Fenske, G.R.; Emerson, J.E.; Savoie, F.E.

    1983-12-01

    The gas release behavior of irradiated EBR-II fuel was observed to be dependent on several factors: the presence of cladding, the retained gas content, and the energy absorbed. Fuel that retained in excess of 16 to 17 μmoles/g of fission gas underwent spallation as the cladding melted and released 22 to 45% of its retained gas, while fuel with retained gas levels below approx. 15 to 16 μmoles/g released less than approx. 9% of its gas as the cladding melted. During subsequent direct electrical heating ramps, fuel that did not spall released an additional quantity of gas (up to 4 μmoles/g), depending on the energy absorbed

  13. Flow-induced and acoustically induced vibration experience in operating gas-cooled reactors

    International Nuclear Information System (INIS)

    Halvers, L.J.

    1977-03-01

    An overview has been presented of flow-induced and acoustically induced vibration failures that occurred in the past in gas-cooled graphite-moderated reactors, and the importance of this experience for the Gas-Cooled Fast-Breeder Reactor (GCFR) project has been assessed. Until now only failures in CO 2 -cooled reactors have been found. No problems with helium-cooled reactors have been encountered so far. It is shown that most of the failures occurred because flow-induced and acoustically induced dynamic loads were underestimated, while at the same time not enough was known about the influence of environmental parameters on material behavior. All problems encountered were solved. The comparison of the influence of the gas properties on acoustically induced and flow-induced vibration phenomena shows that the interaction between reactor design and the thermodynamic properties of the primary coolant precludes a general preference for either carbon dioxide or helium. The acoustic characteristics of CO 2 and He systems are different, but the difference in dynamic loadings due to the use of one rather than the other remains difficult to predict. A slight preference for helium seems, however, to be justified

  14. Preliminary results of Resistive Plate Chambers operated with eco-friendly gas mixtures for application in the CMS experiment

    CERN Document Server

    Abbrescia, Marcello; Benussi, Luigi; Bianco, Stefano; Cauwenbergh, Simon Marc D; Ferrini, Mauro; Muhammad, Saleh; Passamontic, L; Pierluigi, Daniele; Piccolo, Davide; Primavera, Federica; Russo, Alessandro; Savianoc, G; Tytgat, Michael

    2016-01-01

    The operations of Resistive Plate Chambers in LHC experiments require F-based gases for optimal performance. Recent regulations demand the use of environmentally unfriendly F-based gases to be limited or banned. In view of the CMS experiment upgrade several tests are ongoing to measure the performance of the detector in terms of efficiency, streamer probability, induced charge and time resolution. Prototype chambers with readout pads and with the standard cms electronic setup are under test. In this talk preliminary results on performance of RPCs operated with a potential eco-friendly gas candidate 1,3,3,3-Tetrafluoropropene, commercially known as HFO-1234ze and with CO2 based gas mixtures are presented and discussed for the possible application in the CMS experiment.

  15. Experiment data report for Semiscale Mod-1 Test S-05-2 (alternate ECC injection test)

    International Nuclear Information System (INIS)

    Feldman, E.M.; Collins, B.L.; Sackett, K.E.

    1977-02-01

    Recorded test data are presented for Test S-05-2 of the Semiscale Mod-1 alternate emergency core coolant (ECC) injection test series. This test is one of several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-2 was conducted from an initial cold leg fluid temperature of 545 0 F and an initial pressure of 2263 psia. A simulated double-ended offset shear cold leg break was used to investigate core and system response to a depressurization and reflood transient with ECC injection at the intact loop pump suction and broken loop cold leg. A reduced lower plenum volume was used for this test to more accurately represent the lower plenum of a PWR, based on system volume scaling. System flow was set to achieve a core fluid temperature differential of 65 0 F at a core power level of 1.44 MW. The flow resistance of the intact loop was based on core area scaling. An electrically heated core with a slightly peaked radial power profile was used in the pressure vessel to simulate the predicted surface heat flux of nuclear fuel rods during a loss-of-coolant accident

  16. Corporate social policy - problems of institutionalization and experience of Russian oil and gas companies

    Science.gov (United States)

    Nekhoda, E.; Kolbysheva, Yu; Makoveeva, V.

    2015-11-01

    The article examines a range of problems related to the process of institutionalization in the corporate social policy, characterizing the social responsibility of business and representing a part of the general strategy of corporate social responsibility. The experience of the social policy implementation in oil and gas companies is analyzed.

  17. Recent experience with onshore oil and gas operations in the Mackenzie Delta, NWT

    International Nuclear Information System (INIS)

    Burns, J.

    1999-01-01

    Hydrocarbon deposits in the Beaufort Sea and Mackenzie Delta indicate mean discovered gas reserves of 5 trillion cubic feet of natural gas, 67 million barrels of condensate, and 247 million barrels of oil in fields located onshore. There may be even bigger undiscovered reserves that could be proven by a surge in drilling likely to occur in this region within the next few years. There are a number of characteristics of this area that appeal to the oil and gas industry over and above the discovered and undiscovered reserves. There is a local aboriginal group with a settled land claim, clear and reasonable rules for access, a business-like approach to development and a sophisicated understanding of the oil and gas industry. There is reasonable access by road, commercial air service, rail and barge by Hay river or sea with an excellent harbour at Tuktoyaktuk. Local contractors and labour with applicable skills and good equipment are available. The Inuvialuit Petroleum Corp. and its partners Altagas Services Inc. and Enbridge Inc. completed a project to supply the town of Inuvik with natural gas for electricity generation and local distribution. This project is a small example of the physical, economic and regulatory environments that the oil industry will face with the undertaking of larger projects. Aspects of the region described include: the Inuvialuit, recent experience, logistics, regulatory environment, project approvels, environmental, and specific observations

  18. Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

    International Nuclear Information System (INIS)

    Kamide, H.; Ieda, Y.; Toda, S.; Isozaki, T.; Sugawara, S.

    1993-01-01

    Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor core during natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

  19. Multi-bundle sodium experiments for thermohydraulics in core subassemblies during natural circulation decay heat removal operation

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, H; Ieda, Y; Toda, S; Isozaki, T; Sugawara, S [Reactor Engineering Section, O-arai Engineering Center, Power Reactor and Nuclear Fuel Development Corporation, Narita, O-arai, Ibaraki-ken (Japan)

    1993-02-01

    Two types of multi-subassembly sodium experiments, CCTL-CFR tests and PLANDTL-DHX tests, have been carried out in order to investigate thermohydraulics in a fast reactor coreduring natural circulation. Basic experiments are carried out in CCTL-CFR test rig without inter-wrapper gap and under steady state. Integral experiments are performed in PLANDTL-DHX test rig with the inter-wrapper gap and a dip cooler in an upper plenum under steady state and transient conditions. The first series of the experiments and post analyses showed that inter-subassembly heat transfer had significant effects on the transverse temperature distribution in the subassembly and was strongly coupled with intra-subassembly flow redistribution. And the cold sodium provided by the dip cooler could reduce the hot spot temperature in the pin bundle mainly via the inter-wrapper gap. (author)

  20. FBR type reactors

    International Nuclear Information System (INIS)

    Otsuka, Masaya; Yamakawa, Masanori; Goto, Tadashi; Ikeuchi, Toshiaki; Yamaki, Hideo.

    1986-01-01

    Purpose: To prevent thermal deformation and making the container compact by improving the cooling performance of main container walls. Constitution: A pipeway is extended from a high pressure plenum below the reactor core and connected to the lower side of the flow channel at the inside of a thermal shielding layer disposed to the inside of the main container wall. Low pressure sodium sent from the low temperature plenum into the high pressure plenum is introduced to the pipeway, caused to uprise in the inside flow channel, then turned for the direction, caused to descend in the outer side flow channel between the main container and the inside flow channel and then returned to the low temperature plenum. A heat insulating layer disposed with argon gas is installed to the inside of the flow channel to reduce the temperature change applied upon reactor scram. An annular linear induction pump capable of changing the voltage polarity is disposed at the midway of the pipeway and the polarity is switched such that the direction of flow of the liquid sodium is exerted as a braking force upon rated operation, whereas exerted as a pumping force upon reactor scram. (Sekiya, K.)

  1. Radon mitigation experience in difficult-to-mitigate schools

    International Nuclear Information System (INIS)

    Leovic, K.W.; Craig, A.B.

    1990-01-01

    Initial radon mitigation experience in schools has shown sub-slab depressurization (SSD) to be generally effective in reducing elevated levels of radon in schools that have a continuous layer of clean, coarse aggregate underneath the slab. However, mitigation experience is limited in schools without sub-slab aggregate and in schools with characteristics such as return-air ductwork underneath the slab or unducted return-air plenums in the drop ceiling that are open to the sub-slab area (via open tops of block walls). Mitigation of schools with utility tunnels and of schools constructed over crawl spaces is also limited. Three Maryland schools exhibiting some of the above characteristics are being researched to help understand the mechanisms that control radon entry and mitigation in schools where standard SSD systems are not effective. This paper discusses specific characteristics of potentially difficult-to-mitigate schools and, where applicable, details examples from the three Maryland schools

  2. Irradiation induced aerosol formation in flue gas: experiments on low doses

    International Nuclear Information System (INIS)

    Maekelae, J.M.

    1992-01-01

    Laboratory experiments on irradiation induced aerosol formation from gaseous sulphur dioxide in humid air are presented. This work is connected to the aerosol particle formation process in the electron beam technique for cleaning flue gas. As a partial process of this method primary products of the radiolysis of water vapour convert sulphur dioxide into gaseous sulphuric acid which then nucleates with water vapour forming small acid droplets. This experimental work has been performed on relatively low absorbed doses. Aerosol particle formation is strongly dependent on dose. In the experiments, the first aerosol particles were detected already on absorbed doses of 0.1-10 mGy. The particle size in these cases is in the so-called ultrafine size range (1-20 nm). In this article three experimental set-ups with some characteristic results are presented. (Author)

  3. Gas transport during in vitro and in vivo preclinical testing of inert gas therapies

    Directory of Open Access Journals (Sweden)

    Ira Katz

    2016-01-01

    Full Text Available New gas therapies using inert gases such as xenon and argon are being studied, which require in vitro and in vivo preclinical experiments. Examples of the kinetics of gas transport during such experiments are analyzed in this paper. Using analytical and numerical models, we analyze an in vitro experiment for gas transport to a 96 cell well plate and an in vivo delivery to a small animal chamber, where the key processes considered are the wash-in of test gas into an apparatus dead volume, the diffusion of test gas through the liquid media in a well of a cell test plate, and the pharmacokinetics in a rat. In the case of small animals in a chamber, the key variable controlling the kinetics is the chamber wash-in time constant that is a function of the chamber volume and the gas flow rate. For cells covered by a liquid media the diffusion of gas through the liquid media is the dominant mechanism, such that liquid depth and the gas diffusion constant are the key parameters. The key message from these analyses is that the transport of gas during preclinical experiments can be important in determining the true dose as experienced at the site of action in an animal or to a cell.

  4. Numerical modeling of plasma plume evolution against ambient background gas in laser blow off experiments

    International Nuclear Information System (INIS)

    Patel, Bhavesh G.; Das, Amita; Kaw, Predhiman; Singh, Rajesh; Kumar, Ajai

    2012-01-01

    Two dimensional numerical modelling based on simplified hydrodynamic evolution for an expanding plasma plume (created by laser blow off) against an ambient background gas has been carried out. A comparison with experimental observations shows that these simulations capture most features of the plasma plume expansion. The plume location and other gross features are reproduced as per the experimental observation in quantitative detail. The plume shape evolution and its dependence on the ambient background gas are in good qualitative agreement with the experiment. This suggests that a simplified hydrodynamic expansion model is adequate for the description of plasma plume expansion.

  5. A summary of computational experience at GE Aircraft Engines for complex turbulent flows in gas turbines

    Science.gov (United States)

    Zerkle, Ronald D.; Prakash, Chander

    1995-01-01

    This viewgraph presentation summarizes some CFD experience at GE Aircraft Engines for flows in the primary gaspath of a gas turbine engine and in turbine blade cooling passages. It is concluded that application of the standard k-epsilon turbulence model with wall functions is not adequate for accurate CFD simulation of aerodynamic performance and heat transfer in the primary gas path of a gas turbine engine. New models are required in the near-wall region which include more physics than wall functions. The two-layer modeling approach appears attractive because of its computational complexity. In addition, improved CFD simulation of film cooling and turbine blade internal cooling passages will require anisotropic turbulence models. New turbulence models must be practical in order to have a significant impact on the engine design process. A coordinated turbulence modeling effort between NASA centers would be beneficial to the gas turbine industry.

  6. Gas hydrate nucleation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    The overall aim of the project was to gain more knowledge about the kinetics of gas hydrate formation especially the early growth phase. Knowledge of kinetics of gas hydrate formation is important and measurements of gas hydrate particle size and concentration can contribute to improve this knowledge. An experimental setup for carrying out experimental studies of the nucleation and growth of gas hydrates has been constructed and tested. Multi wavelength extinction (MWE) was the experimental technique selected for obtaining particle diameter and concentration. The principle behind MWE is described as well as turbidity spectrum analysis that in an initial stage of the project was considered as an alternative experimental technique. Details of the experimental setup and its operation are outlined. The measuring cell consists of a 1 litre horizontal tube sustaining pressures up to 200 bar. Laser light for particle size determination can be applied through sapphire windows. A description of the various auxiliary equipment and of another gas hydrate cell used in the study are given. A computer program for simulation and analysis of gas hydrate experiments is based on the gas hydrate kinetics model proposed by Skovborg and Rasmussen (1993). Initial measurements showed that knowledge of the refractive index of gas hydrates was important in order to use MWE. An experimental determination of the refractive index of methane and natural gas hydrate is described. The test experiments performed with MWE on collectives of gas hydrate particles and experiments with ethane, methane and natural gas hydrate are discussed. Gas hydrate particles initially seem to grow mainly in size and at latter stages in number. (EG) EFP-94; 41 refs.

  7. Gas hydrate nucleation

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-31

    The overall aim of the project was to gain more knowledge about the kinetics of gas hydrate formation especially the early growth phase. Knowledge of kinetics of gas hydrate formation is important and measurements of gas hydrate particle size and concentration can contribute to improve this knowledge. An experimental setup for carrying out experimental studies of the nucleation and growth of gas hydrates has been constructed and tested. Multi wavelength extinction (MWE) was the experimental technique selected for obtaining particle diameter and concentration. The principle behind MWE is described as well as turbidity spectrum analysis that in an initial stage of the project was considered as an alternative experimental technique. Details of the experimental setup and its operation are outlined. The measuring cell consists of a 1 litre horizontal tube sustaining pressures up to 200 bar. Laser light for particle size determination can be applied through sapphire windows. A description of the various auxiliary equipment and of another gas hydrate cell used in the study are given. A computer program for simulation and analysis of gas hydrate experiments is based on the gas hydrate kinetics model proposed by Skovborg and Rasmussen (1993). Initial measurements showed that knowledge of the refractive index of gas hydrates was important in order to use MWE. An experimental determination of the refractive index of methane and natural gas hydrate is described. The test experiments performed with MWE on collectives of gas hydrate particles and experiments with ethane, methane and natural gas hydrate are discussed. Gas hydrate particles initially seem to grow mainly in size and at latter stages in number. (EG) EFP-94; 41 refs.

  8. Design and operational experience with the off-gas cleaning system of the Seibersdorf incinerator plant

    International Nuclear Information System (INIS)

    Patek, P.

    1982-05-01

    After a description of the design and the construction principles of the incinerator building, the furnace and its attached auxilary devices are explained. The incinerator is layed out for low level wastes. It has a vertical furnace, operates with discontinuous feeding for trashes with heat-values between 600 and 10000 kcal/kg waste. The maximum throughput ammounts 40 kg/h. The purification of the off-gas is guaranteed by a multistage filter system: 2 stages with ceramic candles, an electrostatic filter and a HEPA-filter system. The control of the off-gas cleaning is carried out by a stack instrumentation, consisting of an aerosol-, gas-, iodine- and tritium-monitor; the building is surveilled by doserate- and aerosolmonitors. Finally the experiences of the first year of operation and the main problems in running the plant are described. (Author) [de

  9. Design and operational experience with the off-gas cleaning system of the Seibersdorf incinerator plant

    International Nuclear Information System (INIS)

    Patek, P.R.M.

    1983-01-01

    After a description of the design and the construction principles of the incinerator building, the furnace and its attached auxiliary devices are explained. The incinerator is layed out for low level wastes. It has a vertical furnace, operates with discontinuous feeding for trashes with heat-values between 600 and 10,000 kcal/kg waste. The maximum throughput amounts to 40 kg/h. The purification of the off-gas is guaranteed by a multistage filter system: 2 stages with ceramic candles, an electrostatic filter and a HEPA-filter system. The control of the off-gas cleaning is carried out by a stack instrumentation, consisting of an aerosol-, gas-, iodine- and tritium-monitor; the building is surveyed by doserate and aerosolmonitors. Finally the experiences of the first year of operation and the main problems in running the plant are described. (author)

  10. Measuring Gas-Phase Basicities of Amino Acids Using an Ion Trap Mass Spectrometer: A Physical Chemistry Laboratory Experiment

    Science.gov (United States)

    Sunderlin, Lee S.; Ryzhov, Victor; Keller, Lanea M. M.; Gaillard, Elizabeth R.

    2005-01-01

    An experiment is performed to measure the relative gas-phase basicities of a series of five amino acids to compare the results to literature values. The experiments use the kinetic method for deriving ion thermochemistry and allow students to perform accurate measurements of thermodynamics in a relatively short time.

  11. Track 5: safety in engineering, construction, operations, and maintenance. Reactor physics design, validation, and operating experience. 5. A Negative Reactivity Feedback Device for Actinide Burner Cores

    International Nuclear Information System (INIS)

    Driscoll, M.J.; Hejzlar, P.

    2001-01-01

    Lead-bismuth eutectic (LBE) cooled reactors are of considerable interest because they may be useful for destruction of actinides in a cost-effective manner, particularly cores fueled predominantly with minor actinides, which gain reactivity with burnup. However, they also pose several design challenges: 1. a small (and perhaps even slightly positive) Doppler feedback; 2. small effective delayed neutron yield; 3. a small negative feedback from axial fuel expansion; 4. positive coolant void and temperature coefficients for conventional designs. This has motivated a search for palliative measures, leading to conceptualization of the reactivity feedback device (RFD). The RFD consists of an in-core flask containing helium gas, tungsten wool, and a small reservoir of LBE that communicates with vertical tubes housing neutron absorber floats. The upper part of these guide tubes contains helium gas that is vented into a separate, cooler ex-core helium gas plenum. The principle of operation is as follows: 1. The tungsten wool, hence the helium gas in the in-core plenum, is heated by gammas and loses heat to the walls by convection and conduction (radiation is feeble for monatomic gases and, in any event, intercepted by the tungsten wool). An energy balance determines the gas temperature, hence, pressure, which is 10 atm here. The energy loss rate can be adjusted by using xenon or a gas mixture in place of helium. The tungsten wool mass, which is 1 vol% wool here, can also be increased to increase gamma heating and further retard convection; alternatively, a Dewar flask could be used in place of the additional wool. 2. An increase in core power causes a virtually instantaneous increase in gamma flux, hence, gas heatup: The thermal time constant of the tungsten filaments and their surrounding gas film is ∼40 μs. 3. The increased gas temperature is associated with an increased gas pressure, which forces more liquid metal into the float guide tubes: LBE will rise ∼100 cm

  12. Natural gas extraction and artificial gas injection experiments in Opalinus Clay, Mont Terri rock laboratory (Switzerland)

    Energy Technology Data Exchange (ETDEWEB)

    Vinsot, A.; Lundy, M. [Agence Nationale pour la Gestion des Déchets Radioactifs ANDRA, Meuse Haute-Marne Center, Bure (France); Appelo, C.A.J. [Dr C.A.J. Appelo, Hydrochemical Consultant, Amsterdam (Netherlands); and others

    2017-04-15

    Two experiments have been installed at Mont Terri in 2004 and 2009 that allowed gas circulation within a borehole at a pressure between 1 and 2 bar. These experiments made it possible to observe the natural gases that were initially dissolved in pore-water degassing into the borehole and to monitor their content evolution in the borehole over several years. They also allowed for inert (He, Ne) and reactive (H{sub 2}) gases to be injected into the borehole with the aim either to determine their diffusion properties into the rock pore-water or to evaluate their removal reaction kinetics. The natural gases identified were CO{sub 2}, light alkanes, He, and more importantly N{sub 2}. The natural concentration of four gases in Opalinus Clay pore-water was evaluated at the experiment location: N{sub 2} 2.2 mmol/L ± 25%, CH{sub 4} 0.30 mmol/L ± 25%, C{sub 2}H{sub 6} 0.023 mmol/L ± 25%, C{sub 3}H{sub 8} 0.012 mmol/L ± 25%. Retention properties of methane, ethane, and propane were estimated. Ne injection tests helped to characterize rock diffusion properties regarding the dissolved inert gases. These experimental results are highly relevant towards evaluating how the fluid composition could possibly evolve in the drifts of a radioactive waste disposal facility. (authors)

  13. Experiences with a new soil gas technique for detecting petroleum pollution

    International Nuclear Information System (INIS)

    Mazac, O.; Landa, I.; Rohde, J.R.; Kelly, W.E.

    1996-01-01

    This paper presents field experiences obtained with a new technology for detecting petroleum pollution in soil and ground water based on in situ determination of hydrocarbon concentrations in soil air. Ecoprobe is a new soil gas device from RS-Dynamics in the Czech Republic. The rugged waterproof device is equipped with a built-in computer-controlled semiconductor sensor. Three case histories are presented that demonstrate the use of the equipment under typical conditions. Two case histories present the use of the device under typical field conditions; the third case history compares results from the Ecoprobe and a commercial photoionization detector (PID) device

  14. Study of impact of the AP1000{sup Registered-Sign} reactor vessel upper internals design on fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Xu Yiban; Conner, Michael; Yuan Kun; Dzodzo, Milorad B.; Karoutas, Zeses; Beltz, Steven A.; Ray, Sumit; Bissett, Teresa A. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China)

    2012-11-15

    information from current Westinghouse plants will be shown to demonstrate an experience base of acceptable core lateral flows. From this experience base and the AP1000 CFD results, acceptability of the AP1000 upper plenum design on the fuel performance of the AP1000 fuel design will be demonstrated.

  15. Nuclear fuel element

    International Nuclear Information System (INIS)

    Penrose, R.T.; Thompson, J.R.

    1976-01-01

    A method of protecting the cladding of a nuclear fuel element from internal attack and a nuclear fuel element for use in the core of a nuclear reactor are disclosed. The nuclear fuel element has disposed therein an additive of a barium-containing material and the barium-containing material collects reactive gases through chemical reaction or adsorption at temperatures ranging from room temperature up to fuel element plenum temperatures. The additive is located in the plenum of the fuel element and preferably in the form of particles in a hollow container having a multiplicity of gas permeable openings in one portion of the container with the openings being of a size smaller than the size of the particles. The openings permit gases and liquids entering the plenum to contact the particles. The additive is comprised of elemental barium or a barium alloy containing one or more metals in addition to barium such as aluminum, zirconium, nickel, titanium and combinations thereof. 6 claims, 3 drawing figures

  16. Heat-flow patterns in Tian-Calvet microcalorimeters: Conductive, convective, and radiative transport in gas dosing experiments

    International Nuclear Information System (INIS)

    Vilchiz, Luis Enrique; Pacheco-Vega, Arturo; Handy, Brent E.

    2005-01-01

    Mathematical models of a Tian-Calvet microcalorimeter were solved numerically by the finite-element method in an effort to understand the relative importance of the three basic heat transfer mechanisms operative during gas dosing experiments typically used to determine heats of adsorption on catalysts and adsorbents. The analysis pays particular attention to the quantitative release of heat through various elements of the cell and sensor cups to assess time delays and the deg.ree of thermal shunting that may result in inaccuracies in calorimetric measurements. Conductive transfer predominates in situations where there is high gas headspace pressure. The convection currents that arise when dosing with considerable gas pressure in the cell headspace region are not sufficiently strong to shunt significant amounts of sample heat away from being sensed by the surrounding thermopiles. Therefore, the heat capture fraction (heat sensed/heat produced) does not vary significantly with gas headspace pressure. During gas dosing under very low gas headspace pressure, radiation losses from the top of the sample bed may significantly affect the heat capture fraction, leading to underestimations of adsorption heats, unless the heat radiated from the top of the catalyst bed is effectively reflected back to the sample region or absorbed by an inert packing layer also in thermal contact with the thermopile wall

  17. Fireworks in noble gas clusters a first experiment with the new "free-electron laser"

    CERN Document Server

    2002-01-01

    An international group of scientists has published first experiments carried out using the new soft X-ray free-electron laser (FEL) at the research center DESY in Hamburg, Germany. Using small clusters of noble gas atoms, for the first time, researchers studied the interaction of matter with intense X-ray radiation from an FEL on extremely short time scales (1 page).

  18. R & D of a Gas-Filled RF Beam Profile Monitor for Intense Neutrino Beam Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Yonehara, K. [Fermilab; Backfish, M. [Fermilab; Moretti, A. [Fermilab; Tollestrup, A. V. [Fermilab; Watts, A. [Fermilab; Zwaska, R. M. [Fermilab; Abrams, R. [MUONS Inc., Batavia; Cummings, M. A.; Dudas, A. [MUONS Inc., Batavia; Johnson, R. P. [MUONS Inc., Batavia; Kazakevich, G. [MUONS Inc., Batavia; Neubauer, M. [MUONS Inc., Batavia; Liu, Q. [Case Western Reserve U.

    2017-05-01

    We report the R&D of a novel radiation-robust hadron beam profile monitor based on a gas-filled RF cavity for intense neutrino beam experiments. An equivalent RF circuit model was made and simulated to optimize the RF parameter in a wide beam intensity range. As a result, the maximum acceptable beam intensity in the monitor is significantly increased by using a low-quality factor RF cavity. The plan for the demonstration test is set up to prepare for future neutrino beam experiments.

  19. Utilization of ''CONTACT'' experiments to improve the fission gas release knowledge in PWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Charles, M; Abassin, J J; Bruet, M; Baron, D; Melin, P

    1983-03-01

    The CONTACT experiments, which were carried out by the French CEA, within the framework of a CEA-FRAMATOME collaboration agreement, bear on the behaviour of in-pile irradiated PWR fuel rods. We will focus here upon their results dealing with fission gas release. The experimental device is briefly described, then the following results are given: the kinetics of stable fission gas release for various linear ratings; the instantaneous fractional release rates of radioactive gases versus their decay constant in the range 1.5 10/sup -6/-3.6 10/sup -3/s/sup -1/, for various burnups, as also the influence of fuel temperature. Moreover, the influence of the nature and the pressure of the filling gas upon the release is presented for various linear ratings. The experimental results are discussed and analysed with the purpose to model various physical phenomena involved in the release (low-temperature mechanisms, diffusion).

  20. Utilization of ''CONTACT'' experiments to improve the fission gas release knowledge in PWR fuel rods

    International Nuclear Information System (INIS)

    Charles, M.; Abassin, J.J.; Bruet, M.

    1983-01-01

    The CONTACT experiments, which were carried out by the French CEA, within the framework of a CEA-FRAMATOME collaboration agreement, bear on the behaviour of in-pile irradiated PWR fuel rods. We will focus here upon their results dealing with fission gas release. The experimental device is briefly described, then the following results are given: the kinetics of stable fission gas release for various linear ratings; the instantaneous fractional release rates of radioactive gases versus their decay constant in the range 1.5 10 -6 -3.6 10 -3 s -1 , for various burnups, as also the influence of fuel temperature. Moreover, the influence of the nature and the pressure of the filling gas upon the release is presented for various linear ratings. The experimental results are discussed and analysed with the purpose to model various physical phenomena involved in the release (low-temperature mechanisms, diffusion)

  1. Results of gas exposure experiments for determination of HF concentrations injurious to plants

    Energy Technology Data Exchange (ETDEWEB)

    Guderian, R

    1971-01-01

    Gas exposure experiments were performed under greenhouse conditions to determine the effects of hydrogen fluoride on the growth capacity, yield and quality of plants. Damage to plants was assessed after HF concentrations of 0.85-25 ..mu..g/m/sup 3/. The effects of definite HF quantities on plants are described and relative sensitivities of 17 deciduous trees, 9 evergreens, 24 agricultural garden plants and 17 ornamental plants are presented. 2 references, 7 tables.

  2. Gas-partitioning tracer test to qualify trapped gas during recharge

    Science.gov (United States)

    Heilweil, Victor M.; Kip, Solomon D.; Perkins, Kim S.; Ellett, Kevin M.

    2004-01-01

    Dissolved helium and bromide tracers were used to evaluate trapped gas during an infiltration pond experiment. Dissolved helium preferentially partitioned into trapped gas bubbles, or other pore air, because of its low solubility in water. This produced observed helium retardation factors of as much as 12 relative to bromide. Numerical simulations of helium breakthrough with both equilibrium and kinetically limited advection/dispersion/retardation did not match observed helium concentrations. However, better fits were obtained by including a decay term representing the diffusive loss of helium through interconnected, gas-filled pores. Calculations indicate that 7% to more than 26% of the porosity beneath the pond was filled with gas. Measurements of laboratory hydraulic properties indicate that a 10% decrease in saturation would reduce the hydraulic conductivity by at least one order of magnitude in the well-sorted sandstone, but less in the overlying soils. This is consistent with in situ measurements during the experiment, which show steeper hydraulic gradients in sandstone than in soil. Intrinsic permeability of the soil doubled during the first six months of the experiment, likely caused by a combination of dissolution and thermal contraction of trapped gas. Managers of artificial recharge basins may consider minimizing the amount of trapped gas by using wet, rather than dry, tilling to optimize infiltration rates, particularly in well-sorted porous media in which reintroduced trapped gas may cause substantial reductions in permeability. Trapped gas may also inhibit the amount of focused infiltration that occurs naturally during ephemeral flood events along washes and playas.

  3. Gas-partitioning tracer test to quantify trapped gas during recharge

    Science.gov (United States)

    Heilweil, V.M.; Solomon, D.K.; Perkins, K.S.; Ellett, K.M.

    2004-01-01

    Dissolved helium and bromide tracers were used to evaluate trapped gas during an infiltration pond experiment. Dissolved helium preferentially partitioned into trapped gas bubbles, or other pore air, because of its low solubility in water. This produced observed helium retardation factors of as much as 12 relative to bromide. Numerical simulations of helium breakthrough with both equilibrium and kinetically limited advection/dispersion/retardation did not match observed helium concentrations. However, better fits were obtained by including a decay term representing the diffusive loss of helium through interconnected, gas-filled pores. Calculations indicate that 7% to more than 26% of the porosity beneath the pond was filled with gas. Measurements of laboratory hydraulic properties indicate that a 10% decrease in saturation would reduce the hydraulic conductivity by at least one order of magnitude in the well-sorted sandstone, but less in the overlying soils. This is consistent with in situ measurements during the experiment, which show steeper hydraulic gradients in sandstone than in soil. Intrinsic permeability of the soil doubled during the first six months of the experiment, likely caused by a combination of dissolution and thermal contraction of trapped gas. Managers of artificial recharge basins may consider minimizing the amount of trapped gas by using wet, rather than dry, tilling to optimize infiltration rates, particularly in well-sorted porous media in which reintroduced trapped gas may cause substantial reductions in permeability. Trapped gas may also inhibit the amount of focused infiltration that occurs naturally during ephemeral flood events along washes and playas.

  4. Design of Plant Gas Exchange Experiments in a Variable Pressure Growth Chamber

    Science.gov (United States)

    Corey, Kenneth A.

    1996-01-01

    Sustainable human presence in extreme environments such as lunar and martian bases will require bioregenerative components to human life support systems where plants are used for generation of oxygen, food, and water. Reduced atmospheric pressures will be used to minimize mass and engineering requirements. Few studies have assessed the metabolic and developmental responses of plants to reduced pressure and varied oxygen atmospheres. The first tests of hypobaric pressures on plant gas exchange and biomass production at the Johnson Space Center will be initiated in January 1996 in the Variable Pressure Growth Chamber (VPGC), a large, closed plant growth chamber rated for 10.2 psi. Experiments were designed and protocols detailed for two complete growouts each of lettuce and wheat to generate a general database for human life support requirements and to answer questions about plant growth processes in reduced pressure and varied oxygen environments. The central objective of crop growth studies in the VPGC is to determine the influence of reduced pressure and reduced oxygen on the rates of photosynthesis, dark respiration, evapotranspiration and biomass production of lettuce and wheat. Due to the constraint of one experimental unit, internal controls, called pressure transients, will be used to evaluate rates of CO2 uptake, O2 evolution, and H2O generation. Pressure transients will give interpretive power to the results of repeated growouts at both reduced and ambient pressures. Other experiments involve the generation of response functions to partial pressures of O2 and CO2 and to light intensity. Protocol for determining and calculating rates of gas exchange have been detailed. In order to build these databases and implement the necessary treatment combinations in short time periods, specific requirements for gas injections and removals have been defined. A set of system capability checks will include determination of leakage rates conducted prior to the actual crop

  5. Compressed Natural Gas (CNG) Transit Bus Experience Survey: April 2009--April 2010

    Energy Technology Data Exchange (ETDEWEB)

    Adams, R.; Horne, D. B.

    2010-09-01

    This survey was commissioned by the U.S. Department of Energy (DOE) and the National Renewable Energy Laboratory (NREL) to collect and analyze experiential data and information from a cross-section of U.S. transit agencies with varying degrees of compressed natural gas (CNG) bus and station experience. This information will be used to assist DOE and NREL in determining areas of success and areas where further technical or other assistance might be required, and to assist them in focusing on areas judged by the CNG transit community as priority items.

  6. Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

    International Nuclear Information System (INIS)

    Yanagisawa, Kazuaki; Miyanishi, Hideyuki; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi.

    1991-11-01

    Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85 Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/V f or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO 2 fuel, C (counts/h) the radioactivity of 85 Kr at plenum of the tested fuel rod and V f (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

  7. Experiments on gas springs with adjustable stiffness for a free-piston engine; Experimentelle Untersuchungen an Gasfedern mit einstellbarer Kennlinie fuer einen Freikolbenmotor

    Energy Technology Data Exchange (ETDEWEB)

    Pohl, Sven-Erik; Ferrari, Cornelius; Offinger, Stefan [Deutsches Zentrum fuer Luft- und Raumfahrt (DLR), Stuttgart (Germany). Inst. fuer Fahrzeugkonzepte

    2011-04-15

    In this paper two gas spring designs for a free-piston application are experimentally investigated and evaluated. The free-piston concept demands a gas spring that allows adjustable spring characteristics during operation. Spring characteristics can be changed either by introducing gas to the cylinder or by adjusting the cylinder volume. For both concepts, experimental setups are provided. The results of the experiments are thermodynamically investigated and further compared to find the best gas spring design. The mass-variable gas spring design can be shown to be preferable over the volume-variable. (orig.)

  8. Experimental Study of the Twin Turbulent Water Jets Using Laser Doppler Anemometry for Validating Numerical Models

    International Nuclear Information System (INIS)

    Wang Huhu; Lee Saya; Hassan, Yassin A.; Ruggles, Arthur E.

    2014-01-01

    The design of next generation (Gen. IV) high-temperature nuclear reactors including gas-cooled and sodium-cooled ones involves massive numerical works especially the Computational Fluid Dynamics (CFD) simulations. The high cost of large-scale experiments and the inherent uncertainties existing in the turbulent models and wall functions of any CFD codes solving Reynolds-averaged Navier-Stokes (RANS) equations necessitate the high-spacial experimental data sets for benchmarking the simulation results. In Gen. IV conceptual reactors, the high- temperature flows mix in the upper plenum before entering the secondary cooling system. The mixing condition should be accurately estimated and fully understood as it is related to the thermal stresses induced in the upper plenum and the magnitudes of output power oscillations due to any changes of primary coolant temperature. The purpose of this study is to use Laser Doppler Anemometry (LDA) technique to measure the flow field of two submerged parallel jets issuing from two rectangular channels. The LDA data sets can be used to validate the corresponding simulation results. The jets studied in this work were at room temperature. The turbulent characteristics including the distributions of mean velocities, turbulence intensities, Reynolds stresses were studied. Uncertainty analysis was also performed to study the errors involved in this experiment. The experimental results in this work are valid for benchmarking any steady-state numerical simulations using turbulence models to solve RANS equations. (author)

  9. Gas source localization and gas distribution mapping with a micro-drone

    International Nuclear Information System (INIS)

    Neumann, Patrick P.

    2013-01-01

    The objective of this Ph.D. thesis is the development and validation of a VTOL-based (Vertical Take Off and Landing) micro-drone for the measurement of gas concentrations, to locate gas emission sources, and to build gas distribution maps. Gas distribution mapping and localization of a static gas source are complex tasks due to the turbulent nature of gas transport under natural conditions and becomes even more challenging when airborne. This is especially so, when using a VTOL-based micro-drone that induces disturbances through its rotors, which heavily affects gas distribution. Besides the adaptation of a micro-drone for gas concentration measurements, a novel method for the determination of the wind vector in real-time is presented. The on-board sensors for the flight control of the micro-drone provide a basis for the wind vector calculation. Furthermore, robot operating software for controlling the micro-drone autonomously is developed and used to validate the algorithms developed within this Ph.D. thesis in simulations and real-world experiments. Three biologically inspired algorithms for locating gas sources are adapted and developed for use with the micro-drone: the surge-cast algorithm (a variant of the silkworm moth algorithm), the zigzag / dung beetle algorithm, and a newly developed algorithm called ''pseudo gradient algorithm''. The latter extracts from two spatially separated measuring positions the information necessary (concentration gradient and mean wind direction) to follow a gas plume to its emission source. The performance of the algorithms is evaluated in simulations and real-world experiments. The distance overhead and the gas source localization success rate are used as main performance criteria for comparing the algorithms. Next, a new method for gas source localization (GSL) based on a particle filter (PF) is presented. Each particle represents a weighted hypothesis of the gas source position. As a first step, the PF-based GSL algorithm

  10. Gas source localization and gas distribution mapping with a micro-drone

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, Patrick P.

    2013-07-01

    The objective of this Ph.D. thesis is the development and validation of a VTOL-based (Vertical Take Off and Landing) micro-drone for the measurement of gas concentrations, to locate gas emission sources, and to build gas distribution maps. Gas distribution mapping and localization of a static gas source are complex tasks due to the turbulent nature of gas transport under natural conditions and becomes even more challenging when airborne. This is especially so, when using a VTOL-based micro-drone that induces disturbances through its rotors, which heavily affects gas distribution. Besides the adaptation of a micro-drone for gas concentration measurements, a novel method for the determination of the wind vector in real-time is presented. The on-board sensors for the flight control of the micro-drone provide a basis for the wind vector calculation. Furthermore, robot operating software for controlling the micro-drone autonomously is developed and used to validate the algorithms developed within this Ph.D. thesis in simulations and real-world experiments. Three biologically inspired algorithms for locating gas sources are adapted and developed for use with the micro-drone: the surge-cast algorithm (a variant of the silkworm moth algorithm), the zigzag / dung beetle algorithm, and a newly developed algorithm called ''pseudo gradient algorithm''. The latter extracts from two spatially separated measuring positions the information necessary (concentration gradient and mean wind direction) to follow a gas plume to its emission source. The performance of the algorithms is evaluated in simulations and real-world experiments. The distance overhead and the gas source localization success rate are used as main performance criteria for comparing the algorithms. Next, a new method for gas source localization (GSL) based on a particle filter (PF) is presented. Each particle represents a weighted hypothesis of the gas source position. As a first step, the PF

  11. Gas source localization and gas distribution mapping with a micro-drone

    Energy Technology Data Exchange (ETDEWEB)

    Neumann, Patrick P.

    2013-07-01

    The objective of this Ph.D. thesis is the development and validation of a VTOL-based (Vertical Take Off and Landing) micro-drone for the measurement of gas concentrations, to locate gas emission sources, and to build gas distribution maps. Gas distribution mapping and localization of a static gas source are complex tasks due to the turbulent nature of gas transport under natural conditions and becomes even more challenging when airborne. This is especially so, when using a VTOL-based micro-drone that induces disturbances through its rotors, which heavily affects gas distribution. Besides the adaptation of a micro-drone for gas concentration measurements, a novel method for the determination of the wind vector in real-time is presented. The on-board sensors for the flight control of the micro-drone provide a basis for the wind vector calculation. Furthermore, robot operating software for controlling the micro-drone autonomously is developed and used to validate the algorithms developed within this Ph.D. thesis in simulations and real-world experiments. Three biologically inspired algorithms for locating gas sources are adapted and developed for use with the micro-drone: the surge-cast algorithm (a variant of the silkworm moth algorithm), the zigzag / dung beetle algorithm, and a newly developed algorithm called ''pseudo gradient algorithm''. The latter extracts from two spatially separated measuring positions the information necessary (concentration gradient and mean wind direction) to follow a gas plume to its emission source. The performance of the algorithms is evaluated in simulations and real-world experiments. The distance overhead and the gas source localization success rate are used as main performance criteria for comparing the algorithms. Next, a new method for gas source localization (GSL) based on a particle filter (PF) is presented. Each particle represents a weighted hypothesis of the gas source position. As a first step, the PF-based GSL algorithm

  12. First Argon Gas Puff Experiments With 500 ns Implosion Time On Sphinx Driver

    Science.gov (United States)

    Zucchini, F.; Calamy, H.; Lassalle, F.; Loyen, A.; Maury, P.; Grunenwald, J.; Georges, A.; Morell, A.; Bedoch, J.-P.; Ritter, S.; Combes, P.; Smaniotto, O.; Lample, R.; Coleman, P. L.; Krishnan, M.

    2009-01-01

    Experiments have been performed at the SPHINX driver to study potential of an Argon Gas Puff load designed by AASC. We present here the gas Puff hardware and results of the last shot series. The Argon Gas Puff load used is injected thanks to a 20 cm diameter nozzle. The nozzle has two annuli and a central jet. The pressure and gas type in each of the nozzle plena can be independently adjusted to tailor the initial gaz density distribution. This latter is selected as to obtain an increasing radial density from outer shell towards the pinch axis in order to mitigate the RT instabilities and to increase radiating mass on axis. A flashboard unit produces a high intensity UV source to pre-ionize the Argon gas. Typical dimensions of the load are 200 mm in diameter and 40 mm height. Pressures are adjusted to obtain an implosion time around 550 ns with a peak current of 3.5 MA. With the goal of improving k-shell yield a mass scan of the central jet was performed and implosion time, mainly given by outer and middle plena settings, was kept constant. Tests were also done to reduce the implosion time for two configurations of the central jet. Strong zippering of the radiation production was observed mainly due to the divergence of the central jet over the 40 mm of the load height. Due to that feature k-shell radiation is mainly obtained near cathode. Therefore tests were done to mitigate this effect first by adjusting local pressure of middle and central jet and second by shortening the pinch length. At the end of this series, best shot gave 5 kJ of Ar k-shell yield. PCD detectors showed that k-shell x-ray power was 670 GW with a FWHM of less than 10 ns.

  13. First Argon Gas Puff Experiments With 500 ns Implosion Time On Sphinx Driver

    International Nuclear Information System (INIS)

    Zucchini, F.; Calamy, H.; Lassalle, F.; Loyen, A.; Maury, P.; Grunenwald, J.; Georges, A.; Morell, A.; Bedoch, J.-P.; Ritter, S.; Combes, P.; Smaniotto, O.; Lample, R.; Coleman, P. L.; Krishnan, M.

    2009-01-01

    Experiments have been performed at the SPHINX driver to study potential of an Argon Gas Puff load designed by AASC. We present here the gas Puff hardware and results of the last shot series.The Argon Gas Puff load used is injected thanks to a 20 cm diameter nozzle. The nozzle has two annuli and a central jet. The pressure and gas type in each of the nozzle plena can be independently adjusted to tailor the initial gaz density distribution. This latter is selected as to obtain an increasing radial density from outer shell towards the pinch axis in order to mitigate the RT instabilities and to increase radiating mass on axis. A flashboard unit produces a high intensity UV source to pre-ionize the Argon gas. Typical dimensions of the load are 200 mm in diameter and 40 mm height. Pressures are adjusted to obtain an implosion time around 550 ns with a peak current of 3.5 MA.With the goal of improving k-shell yield a mass scan of the central jet was performed and implosion time, mainly given by outer and middle plena settings, was kept constant. Tests were also done to reduce the implosion time for two configurations of the central jet. Strong zippering of the radiation production was observed mainly due to the divergence of the central jet over the 40 mm of the load height. Due to that feature k-shell radiation is mainly obtained near cathode. Therefore tests were done to mitigate this effect first by adjusting local pressure of middle and central jet and second by shortening the pinch length.At the end of this series, best shot gave 5 kJ of Ar k-shell yield. PCD detectors showed that k-shell x-ray power was 670 GW with a FWHM of less than 10 ns.

  14. Evaluation of Gas Retention in Waste Simulants: Tall Column Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Schonewill, Philip P.; Gauglitz, Phillip A.; Shimskey, Rick W.; Denslow, Kayte M.; Powell, Michael R.; Boeringa, Gregory K.; Bontha, Jagannadha R.; Karri, Naveen K.; Fifield, Leonard S.; Tran, Diana N.; Sande, Susan; Heldebrant, David J.; Meacham, Joseph E.; Smet, Dave; Bryan, Wesley E.; Calmus, Ronald B.

    2014-05-16

    Gas generation in Hanford’s underground waste storage tanks can lead to gas accumulation within the layer of settled solids (sludge) at the tank bottom. The gas, which typically has hydrogen as the major component together with other flammable species, is formed principally by radiation-driven chemical reactions. Accumulation of these gases within the sludge in a waste tank is undesirable and limits the amount of tank volume for waste storage. Further, accumulation of large amounts of gas in the sludge may potentially result in an unacceptable release of the accumulated gas if the sludge-layer density is reduced to less than that of the overlying sludge or that of the supernatant liquid. Rapid release of large amounts of flammable gases could endanger personnel and equipment near the tank. For this reason, a thorough understanding of the circumstances that can lead to a potentially problematic gas accumulation in sludge layers is needed. To respond to this need, the Deep Sludge Gas Release Event Program (DSGREP) was commissioned to examine gas release behavior in sludges.

  15. In gas laser ionization and spectroscopy experiments at the Superconducting Separator Spectrometer (S3): Conceptual studies and preliminary design

    International Nuclear Information System (INIS)

    Ferrer, R.; Bastin, B.; Boilley, D.; Creemers, P.; Delahaye, P.; Liénard, E.; Fléchard, X.; Franchoo, S.; Ghys, L.; Huyse, M.; Kudryavtsev, Yu.; Lecesne, N.; Lu, H.; Lutton, F.; Mogilevskiy, E.; Pauwels, D.; Piot, J.; Radulov, D.; Rens, L.; Savajols, H.

    2013-01-01

    Highlights: • A setup to perform In-Gas Laser Ionization and Spectroscopy experiments at the Super Separator Spectrometer is presented. • The reported studies address important aspects necessary to applied the IGLIS technique to short-lived isotopes. • An R and D phase required to reach an enhanced spectral resolution will be carried out at KU Leuven. • High-sensitivity and enhanced-resolution laser spectroscopy studies will be possible with the IGLIS setup at S 3 . -- Abstract: The results of preparatory experiments and the preliminary designs of a new in-gas laser ionization and spectroscopy setup, to be coupled to the Super Separator Spectrometer S 3 of SPIRAL2-GANIL, are reported. Special attention is given to the development and tests to carry out a full implementation of the in-gas jet laser spectroscopy technique. Application of this novel technique to radioactive species will allow high-sensitivity and enhanced-resolution laser spectroscopy studies of ground- and excited-state properties of exotic nuclei

  16. Some experiments on cold fusion by deuterium hydrogen gas infusion in titanium metal alloy

    International Nuclear Information System (INIS)

    Mestnik Filho, J.; Geraldo, L.P.; Pugliese, R.; Saxena, R.N.; Morato, S.P.; Fulfaro, R.

    1990-05-01

    New results on cold fusion are reported where three different experimental situations have been tried: a) deuterium gas loaded titanium; b) deuterium gas loaded Ti 0.8 Zr 0.2 CrMn alloy and c) titanium and the Ti 0.8 Zr 0.2 CrMn alloy loaded with a mixture of deuterium and hydrogen gases. With these experiments, new thermodynamical non equilibrium conditions were achieved and the possibility of cold fusion between protons and deuterons was also tested. Three independent neutron detectors and one NaI(Tl) were utilized. Despite some large values reported in the literature for the fusion rate, an upper limit of only 8 x 10 -24 fusions/sper deuterium pair or per deuterium-hydrogen pair was determined within the attained accuracy. (author) [pt

  17. Laser absorption, power transfer, and radiation symmetry during the first shock of inertial confinement fusion gas-filled hohlraum experiments

    International Nuclear Information System (INIS)

    Pak, A.; Dewald, E. L.; Landen, O. L.; Milovich, J.; Strozzi, D. J.; Berzak Hopkins, L. F.; Bradley, D. K.; Divol, L.; Ho, D. D.; MacKinnon, A. J.; Meezan, N. B.; Michel, P.; Moody, J. D.; Moore, A. S.; Schneider, M. B.; Town, R. P. J.; Hsing, W. W.; Edwards, M. J.

    2015-01-01

    Temporally resolved measurements of the hohlraum radiation flux asymmetry incident onto a bismuth coated surrogate capsule have been made over the first two nanoseconds of ignition relevant laser pulses. Specifically, we study the P2 asymmetry of the incoming flux as a function of cone fraction, defined as the inner-to-total laser beam power ratio, for a variety of hohlraums with different scales and gas fills. This work was performed to understand the relevance of recent experiments, conducted in new reduced-scale neopentane gas filled hohlraums, to full scale helium filled ignition targets. Experimental measurements, matched by 3D view factor calculations, are used to infer differences in symmetry, relative beam absorption, and cross beam energy transfer (CBET), employing an analytic model. Despite differences in hohlraum dimensions and gas fill, as well as in laser beam pointing and power, we find that laser absorption, CBET, and the cone fraction, at which a symmetric flux is achieved, are similar to within 25% between experiments conducted in the reduced and full scale hohlraums. This work demonstrates a close surrogacy in the dynamics during the first shock between reduced-scale and full scale implosion experiments and is an important step in enabling the increased rate of study for physics associated with inertial confinement fusion

  18. Laser absorption, power transfer, and radiation symmetry during the first shock of inertial confinement fusion gas-filled hohlraum experiments

    Science.gov (United States)

    Pak, A.; Dewald, E. L.; Landen, O. L.; Milovich, J.; Strozzi, D. J.; Berzak Hopkins, L. F.; Bradley, D. K.; Divol, L.; Ho, D. D.; MacKinnon, A. J.; Meezan, N. B.; Michel, P.; Moody, J. D.; Moore, A. S.; Schneider, M. B.; Town, R. P. J.; Hsing, W. W.; Edwards, M. J.

    2015-12-01

    Temporally resolved measurements of the hohlraum radiation flux asymmetry incident onto a bismuth coated surrogate capsule have been made over the first two nanoseconds of ignition relevant laser pulses. Specifically, we study the P2 asymmetry of the incoming flux as a function of cone fraction, defined as the inner-to-total laser beam power ratio, for a variety of hohlraums with different scales and gas fills. This work was performed to understand the relevance of recent experiments, conducted in new reduced-scale neopentane gas filled hohlraums, to full scale helium filled ignition targets. Experimental measurements, matched by 3D view factor calculations, are used to infer differences in symmetry, relative beam absorption, and cross beam energy transfer (CBET), employing an analytic model. Despite differences in hohlraum dimensions and gas fill, as well as in laser beam pointing and power, we find that laser absorption, CBET, and the cone fraction, at which a symmetric flux is achieved, are similar to within 25% between experiments conducted in the reduced and full scale hohlraums. This work demonstrates a close surrogacy in the dynamics during the first shock between reduced-scale and full scale implosion experiments and is an important step in enabling the increased rate of study for physics associated with inertial confinement fusion.

  19. Laser absorption, power transfer, and radiation symmetry during the first shock of inertial confinement fusion gas-filled hohlraum experiments

    Energy Technology Data Exchange (ETDEWEB)

    Pak, A.; Dewald, E. L.; Landen, O. L.; Milovich, J.; Strozzi, D. J.; Berzak Hopkins, L. F.; Bradley, D. K.; Divol, L.; Ho, D. D.; MacKinnon, A. J.; Meezan, N. B.; Michel, P.; Moody, J. D.; Moore, A. S.; Schneider, M. B.; Town, R. P. J.; Hsing, W. W.; Edwards, M. J. [Lawrence Livermore National Laboratory, Livermore, California, 94550 (United States)

    2015-12-15

    Temporally resolved measurements of the hohlraum radiation flux asymmetry incident onto a bismuth coated surrogate capsule have been made over the first two nanoseconds of ignition relevant laser pulses. Specifically, we study the P2 asymmetry of the incoming flux as a function of cone fraction, defined as the inner-to-total laser beam power ratio, for a variety of hohlraums with different scales and gas fills. This work was performed to understand the relevance of recent experiments, conducted in new reduced-scale neopentane gas filled hohlraums, to full scale helium filled ignition targets. Experimental measurements, matched by 3D view factor calculations, are used to infer differences in symmetry, relative beam absorption, and cross beam energy transfer (CBET), employing an analytic model. Despite differences in hohlraum dimensions and gas fill, as well as in laser beam pointing and power, we find that laser absorption, CBET, and the cone fraction, at which a symmetric flux is achieved, are similar to within 25% between experiments conducted in the reduced and full scale hohlraums. This work demonstrates a close surrogacy in the dynamics during the first shock between reduced-scale and full scale implosion experiments and is an important step in enabling the increased rate of study for physics associated with inertial confinement fusion.

  20. Corporate realignments in the natural gas industry: does the North American experience foretell the future for the European Union?

    Energy Technology Data Exchange (ETDEWEB)

    Rutledge, I.; Wright, Ph. [Sheffield Univ., Energy Studies Programme (United Kingdom); Wright, Ph. [Montpellier-1 Univ., CREDEN-LASER, 34 (France)

    2000-09-01

    This paper seeks to explore the extent to which the corporate realignments which have occurred in the North American Natural Gas Industry during a now relatively lengthy experience with liberalization involving a large number of players, will be imitated in the future by European Union countries other than the UK (which is of course already long-embarked along the path of Anglo-Saxon liberalization). The paper first of all catalogues the North American experience, drawing on company performance data assembled by the authors over the last decade (Rutledge and Wright, 1993, 1995, 1997, 1999, 2000). Secondly, this empirical exploration gives way to theoretical speculation: are there elements of the North American experience for which explanatory generalizations are possible? Thirdly, these empirical and theoretical insights are employed to identify and explore actual and potential differences in the corporate evolution of the European Union natural gas industry. (authors)

  1. Contributed Review: The novel gas puff targets for laser-matter interaction experiments

    Energy Technology Data Exchange (ETDEWEB)

    Wachulak, Przemyslaw W., E-mail: wachulak@gmail.com [Institute of Optoelectronics, Military University of Technology, Ul. Gen. S. Kaliskiego 2, 00-908 Warsaw (Poland)

    2016-09-15

    Various types of targetry are used nowadays in laser matter interaction experiments. Such targets are characterized using different methods capable of acquiring information about the targets such as density, spatial distribution, and temporal behavior. In this mini-review paper, a particular type of target will be presented. The targets under consideration are gas puff targets of various and novel geometries. Those targets were investigated using extreme ultraviolet (EUV) and soft X-ray (SXR) imaging techniques, such as shadowgraphy, tomography, and pinhole camera imaging. Details about characterization of those targets in the EUV and SXR spectral regions will be presented.

  2. European Experience after The Gas Directive On The Business

    Energy Technology Data Exchange (ETDEWEB)

    Holm, Tore

    1999-07-01

    The Gas Directive came into effect in August 1998. Its main building block is Third Party Access (TPA) to gas transmission and distribution. The transposition into national law in the Member States is in progress as planned. A much higher percentage of the market will be open to competition than the minimum required in the Gas Directive. This presentation asserts that those who have attempted to predict the outcome, the process and the timing of the ongoing development are largely people or organizations with vested interests either way or people who have simply ''transposed'' the UK model into a Continental setting. But there are much more important issues for the European gas industry than the EU liberalization process per se. The presentation discusses what the people in Shell call the Low Oil Price World, then the ''Tokyo'' implementation and then gas fundamentals in Europe. Finally, an attempt is given to see how the political and commercial processes that are already under way may unfold in the future.

  3. Study on CFD approach for gas entrainment phenomenon. Evaluation of applicability of finite element flow analysis

    International Nuclear Information System (INIS)

    Eguchi, Yuzuru

    2005-07-01

    The report is concerned with the evaluation of applicability of numerical modelling methods for the prediction of gas entrainment in an upper plenum of a sodium-cooled fast breeder reactor (FBR). Special attention was paid to applicability of variational multiscale (VMS) modelling in the context of the Finite Element Method. Two flow problems, which were experimentally shown to induce gas entrainment, are solved by a VMS code (MISTRAL). First, computing a benchmark problem of a gas entrainment swirl flow in a cylindrical vessel has led to the following results; (1) the VMS solution is able to resolve the precise vortex core structure more accurately than the non-VMS solution computed by Smart-fem. The circumferential velocity obtained from VMS computation rises almost double in comparison with the non-VMS solution, though it still underestimates the experimental values. (2) the half-value radius of the negative region of the second invariant of velocity gradient matches well between the VMS solution and non-VMS solution. (3) the negative/positive boundary of the second invariant of velocity gradient obtained from the VMS solution is closer to the vortex core radius observed in the experiment than that of the non-VMS solution, though the vortex dip length computed from the VMS result is shorter than the experimental value. Second, computing a benchmark problem of open channel flow with a square pillar and downstream suction pipe has led to the following results; (4) 2Δx-type spatial oscillation was observed due to lack of mesh subdivisions. (5) the distributional profile of the second invariant of velocity gradient is similar to that of the first problem (swirl flow in a cylindrical vessel), characterized by a strong negative region surrounded by a weak positive region. As a possible future plan, it may be necessary to analyze more precisely the features of unsteady vortices obtained in the second benchmark problem and to identify the difference (if any) from the

  4. Measurements of Plasma Expansion due to Background Gas in the Electron Diffusion Gauge Experiment

    International Nuclear Information System (INIS)

    Morrison, Kyle A.; Paul, Stephen F.; Davidson, Ronald C.

    2003-01-01

    The expansion of pure electron plasmas due to collisions with background neutral gas atoms in the Electron Diffusion Gauge (EDG) experiment device is observed. Measurements of plasma expansion with the new, phosphor-screen density diagnostic suggest that the expansion rates measured previously were observed during the plasma's relaxation to quasi-thermal-equilibrium, making it even more remarkable that they scale classically with pressure. Measurements of the on-axis, parallel plasma temperature evolution support the conclusion

  5. The non-proliferation experiment and gas sampling as an on-site inspection activity: A progress report

    International Nuclear Information System (INIS)

    Carrigan, C.R.

    1994-03-01

    The Non-proliferation Experiment (NPE) is contributing to the development of gas sampling methods and models that may be incorporated into future on-site inspection (OSI) activities. Surface gas sampling and analysis, motivated by nuclear test containment studies, have already demonstrated the tendency for the gaseous products of an underground nuclear test to flow hundreds of meters to the surface over periods ranging from days to months. Even in the presence of a uniform sinusoidal pressure variation, there will be a net flow of cavity gas toward the surface. To test this barometric pumping effect at Rainier Mesa, gas bottles containing sulfur hexaflouride and 3 He were added to the pre-detonation cavity for the 1 kt chemical explosives test. Pre-detonation measurements of the background levels of both gases were obtained at selected sites on top of the mesa. The background levels of both tracers were found to be at or below mass spectrographic/gas chromatographic sensitivity thresholds in the parts-per-trillion range. Post-detonation, gas chromatographic analyses of samples taken during barometric pressure lows from the sampling sites on the mesa indicate the presence of significant levels (300--600 ppt) of sulfur hexaflouride. However, mass spectrographic analyses of gas samples taken to date do not show the presence of 3 He. To explain these observations, several possibilities are being explored through additional sampling/analysis and numerical modeling. For the NPE, the detonation point was approximately 400 m beneath the surface of Rainier Mesa and the event did not produce significant fracturing or subsidence on the surface of the mesa. Thus, the NPE may ultimately represent an extreme, but useful example for the application and tuning of cavity gas detection techniques

  6. Experiments about the integrity of BWR relief pipes in postulated radiolysis gas combustion. Scenario No.2. Minor steam leakages without any lowering of the water level

    International Nuclear Information System (INIS)

    Friedrich, A.; Grune, J.; Sempert, K.; Stern, G.; Kuznetsov, M.; Redlinger, R.; Breitung, W.; Franke, T.

    2008-01-01

    The experiments described in this article were performed to study this comprehensive radiolysis gas scenario: - The relief pipe is filled completely with radiolysis gas (2H 2 +O 2 ). - After opening of the S and R valve, the radiolysis gas is compressed adiabatically by the incoming steam without mixing. - Roughly at the point of peak pressure in the relief pipe (20 bar) the radiolysis gas ignites. This dynamic scenario was studied in steady-state model experiments with a test pipe which corresponds to the relief pipes installed in KKP-1 in terms of materials, dimensions, and manufacturing control. The initial conditions and boundary conditions of the experiments were conservative. In the course of the tests, the maximum dynamic strain and the residual plastic deformation of the test pipe were measured via the transient detonation load. The maximum dynamic strain measured was 0.75%, the maximum residual plastic strain reached 0.15%. The pipe suffered no other deformation above and beyond this slight plastic strain. The radiolysis gas detonation was simulated very well numerically. Using the calculated pressure loads in a structural dynamics model also showed good agreement with the measured maximum dynamic pipe strains. In this way, the experimental findings were confirmed theoretically. The experiments and the calculations showed that postulated radiolysis gas reactions during pressure relief cannot jeopardize the integrity of the relief pipe. (orig.)

  7. Cross-border trading and transmission networks. The natural gas experience

    International Nuclear Information System (INIS)

    Zavattoni, G.

    1992-01-01

    The volume and sources of gas imports to Europe and cross-border trading regulations for gas imports into and within Europe are outlined. The import of gas from Algeria to Italy is presented as a case study. It deals with the purchase contract, transmission through Tunisia, the crossing of the Strait of Sicily and transportation within Italy. The EC Transit Directive, the Draft Directive(s) on third party access and the impact of the directives on the gas market are then discussed. (UK)

  8. Perry's bio-gas experience 1995 ASME/EPRI radwaste workshop

    International Nuclear Information System (INIS)

    Schwenk, A.K.

    1995-01-01

    The Perry Power Plant has been in commercial operation for about ten years. Although we didn't know it at the time, we now believe our bio-gas problem may have started about seven years ago. Barnwell discovered we had a bio-gas problem about a year and a half ago. We found out we had a bio-gas problem a few hours later. The history associated with this process at Perry is outlined, and past as well as present efforts to monitor this process are also discussed

  9. The spray characteristic of gas-liquid coaxial swirl injector by experiment

    Directory of Open Access Journals (Sweden)

    Chen Chen

    2017-01-01

    Full Text Available Using the laser phase Doppler particle analyzer (PDPA, the spray characteristics of gas-liquid coaxial swirl injector were studied. The Sauter mean diameter (SMD, axial velocity and size data rate were measured under different gas injecting pressure drop and liquid injecting pressure drop. Comparing to a single liquid injection, SMD with gas presence is obviously improved. So the gas presence has a significant effect on the atomization of the swirl injector. What’s more, the atomization effect of gas-liquid is enhanced with the increasing of the gas pressure drop. Under the constant gas pressure drop, the injector has an optimal liquid pressure drop under which the atomization performance is best.

  10. 40 CFR 63.7765 - What definitions apply to this subpart?

    Science.gov (United States)

    2010-07-01

    ... points and then convey the captured gas stream to a control device or to the atmosphere. A capture system... design: duct intake devices, hoods, enclosures, ductwork, dampers, manifolds, plenums, and fans. Cold box.... Emissions limitation means any emissions limit or operating limit. Exhaust stream means gases emitted from a...

  11. Studies of Deteriorated Heat Transfer in Prismatic Cores Stemming from Irradiation-Induced Geometry Distortion

    International Nuclear Information System (INIS)

    Williams, Brian G.; Schultz, Richard R.; McEligot, Don M.; McCreery, Glenn

    2015-01-01

    A reference design for the Next Generation Nuclear Plant (NGNP) is to use General Atomics Modular High Temperature Gas-cooled Reactor (MHTGR). For such a configuration in normal operation, the helium coolant flow proceeds from the upper plenum to the lower plenum principally through the core coolant channels and the interstitial gaps (bypass flow) that separate the prismatic blocks from one another. Only the core prismatic blocks have coolant channels. The interstitial gaps are present throughout the core, the inner reflector region, and the out reflector region. The bypass flows in a prismatic gas-cooled reactor (GCR) are of potential concern because they reduce the desired flow rates in the coolant channels and, thereby, can increase outlet gas temperatures and maximum fuel temperatures. Consequently, it is appropriate to account for bypass flows in reactor thermal gas dynamic analyses. The objectives of this project include the following: fundamentally understand bypass flow and heat transfer at scaled, undistorted conditions and with geometry distortions; develop improved estimates of associated loss coefficients, surface friction and heat transfer for systems and network codes; and obtain related data for validation of CFD (computational fluid dynamic) or system (e.g., RELAP5) codes which can be employed in predictions for a GCR for normal power, reduced power, and residual heat removal operations.

  12. Studies of Deteriorated Heat Transfer in Prismatic Cores Stemming from Irradiation-Induced Geometry Distortion

    Energy Technology Data Exchange (ETDEWEB)

    Williams, Brian G. [Idaho State Univ., Pocatello, ID (United States); Schultz, Richard R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); McEligot, Don M. [Univ. of Idaho, Moscow, ID (United States); McCreery, Glenn [Battelle Energy Alliance, LLC, Idaho Falls, ID (United States)

    2015-08-31

    A reference design for the Next Generation Nuclear Plant (NGNP) is to use General Atomics Modular High Temperature Gas-cooled Reactor (MHTGR). For such a configuration in normal operation, the helium coolant flow proceeds from the upper plenum to the lower plenum principally through the core coolant channels and the interstitial gaps (bypass flow) that separate the prismatic blocks from one another. Only the core prismatic blocks have coolant channels. The interstitial gaps are present throughout the core, the inner reflector region, and the out reflector region. The bypass flows in a prismatic gas-cooled reactor (GCR) are of potential concern because they reduce the desired flow rates in the coolant channels and, thereby, can increase outlet gas temperatures and maximum fuel temperatures. Consequently, it is appropriate to account for bypass flows in reactor thermal gas dynamic analyses. The objectives of this project include the following: fundamentally understand bypass flow and heat transfer at scaled, undistorted conditions and with geometry distortions; develop improved estimates of associated loss coefficients, surface friction and heat transfer for systems and network codes; and obtain related data for validation of CFD (computational fluid dynamic) or system (e.g., RELAP5) codes which can be employed in predictions for a GCR for normal power, reduced power, and residual heat removal operations.

  13. Gas migration from oil and gas fields and associated hazards

    International Nuclear Information System (INIS)

    Gurevich, A.E.; Endres, B.L.; Robertson Jr, J.O.; Chilingar, G.V.

    1993-01-01

    The migration of gas from oil and gas formations to the surface is a problem that greatly affects those surface areas where human activity exists. Underground gas storage facilities and oil fields have demonstrated a long history of gas migration problems. Experience has shown that the migration of gas to the surface creates a serious potential risk of explosion, fires, noxious odors and potential emissions of carcinogenic chemicals. These risks must be seriously examined for all oil and gas operations located in urban areas. This paper presents the mechanics of gas migration, paths of migration and a review of a few of the risks that should be considered when operating a gas facility in an urban area. The gas can migrate in a continuous or discontinuous stream through porous, water-filled media to the surface. The primary force in this migration of gas is the difference between specific weights of gas and water

  14. Quantitative study of quasi-one-dimensional Bose gas experiments via the stochastic Gross-Pitaevskii equation

    International Nuclear Information System (INIS)

    Cockburn, S. P.; Gallucci, D.; Proukakis, N. P.

    2011-01-01

    The stochastic Gross-Pitaevskii equation is shown to be an excellent model for quasi-one-dimensional Bose gas experiments, accurately reproducing the in situ density profiles recently obtained in the experiments of Trebbia et al.[Phys. Rev. Lett. 97, 250403 (2006)] and van Amerongen et al.[Phys. Rev. Lett. 100, 090402 (2008)] and the density fluctuation data reported by Armijo et al.[Phys. Rev. Lett. 105, 230402 (2010)]. To facilitate such agreement, we propose and implement a quasi-one-dimensional extension to the one-dimensional stochastic Gross-Pitaevskii equation for the low-energy, axial modes, while atoms in excited transverse modes are treated as independent ideal Bose gases.

  15. Generation of low-Btu fuel gas from agricultural residues experiments with a laboratory scale gas producer

    Energy Technology Data Exchange (ETDEWEB)

    Williams, R O

    1977-01-01

    Two successive laboratory-scale, downdraft gas producers were fabricated and tested. Agricultural and food processing residues including walnut shells, corn cobs, tree prunings, and cotton gin waste, were converted to a low Btu producer gas. The performance of 2 spark ignition engines, when running on producer gas, was highly satisfactory. The ability of the producer to maintain a continuous supply of good quality gas was determined largely by firebox configuration. Fuel handling and fuel flow control problems tended to be specific to individual types of residues. During each test run, air input, firebox temperature, fuel consumption rate, and pressure differential across the producer were monitored. An overall conversion efficiency of 65% was achieved.

  16. Prompt neutron decay constant estimation of RSG-GAS at high power noise experiment

    International Nuclear Information System (INIS)

    Jujuratisbela, U.; Kristedjo; Tukiran; Pinem, S.; Iman, J.; Puryono; Sanjaya, A.; Suwarno

    1998-01-01

    The determination of prompt neutron decay constant (α) of RGS-GAS by using low power noise experiment method at the equilibrium core indicated that the result is not good. The bad result was due to the small ratio of the noise signal to background which was caused by low detector efficiency or contaminated core after long time operation. To solve the problem is tried by using noise experiment technique at high power. The voltage output of neutron detectors at power of 5, 12, and 23 MW were connected to preamplifier and filter then to the Dynamic Signal Analyzer Version-2 and then the power spectral density of each channel of JKT04 and JKT03, the cut off frequency of each channel can be determined by using linear regression technique such that the prompt neutron decay constant can be estimated

  17. Experimental study of core thermohydraulics in fast reactors during transition from forced to natural circulation. Influence of inter-wrapper flow

    International Nuclear Information System (INIS)

    Kamide, H.; Hayashi, K.; Momoi, K.

    1997-01-01

    The evaluation of core thermohydraulics under natural circulation conditions is important to utilize inherent safety features of Fast Reactors. When heat exchangers of a decay heat removal system are operated in an upper plenum of reactor vessel, cold sodium is provided by the heat exchangers. Core-plenum interactions will occur during a natural circulation condition due to this cold sodium in the upper plenum, e.g., it can penetrate into gap regions between fuel subassemblies (inter-wrapper flow, IWF) and the flow may reverse in low power core channels. These interactions will significantly modify the flow and temperature distributions in the core. Sodium experiments were carried out to study these phenomena. In a test section, seven subassemblies are housed and connected to an upper plenum. The influences of core-plenum interactions on the core thermohydraulics were investigated under steady state conditions and also in transitions from forced to natural circulation. Cooling effects of IWF on the fuel subassemblies were found in spite of natural circulation flow reduction in the primary loop due to temperature decreases in the upper non-heated section in the core. The inter-wrapper flow can effectively cool the core under extreme conditions of low flow rates through the core. (author)

  18. Posttest REALP4 analysis of LOFT experiment L1-3A

    International Nuclear Information System (INIS)

    White, J.R.; Holmstrom, H.L.O.

    1977-10-01

    This report presents selected results of posttest RELAP4 modeling of LOFT loss-of-coolant experiment L1-3A, a double-ended isothermal cold leg break with lower plenum emergency core coolant injection. Comparisons are presented between the pretest prediction, the posttest analysis, and the experimental data. It is concluded that pressurizer modeling is important for accurately predicting system behavior during the initial portion of saturated blowdown. Using measured initial conditions rather than nominal specified initial conditions did not influence the system model results significantly. Using finer nodalization in the reactor vessel improved the prediction of the system pressure history by minimizing steam condensation effects. Unequal steam condensation between the downcomer and core volumes appear to cause the manometer oscillations observed in both the pretest and posttest RELAP4 analysis

  19. Loss-of-coolant accident test series TC-1 experiment operating specifications

    International Nuclear Information System (INIS)

    Yackle, T.R.

    1979-09-01

    The purpose of this document is to specify the experiment operating procedure for the test series TC-1. The effects of externally mounted cladding thermocouples on the fuel rod thermal behavior during LOCA blowdown and reflood cycles will be investigated in the test. Potential thermocouple effects include: (a) delayed DNB, (b) momentary cladding rewets following DNB, (c) premature cladding rewet during a blowdown two-phase slug period, and (d) early cladding rewet during reflood. The two-phase slug period will be controlled by momentarily opening the hot leg valve. The slug will consist of lower plenum liquid that is sent through the flow shrouds and will be designed to quench the fuel rods at a rate that is similar to the slug experienced early in the LOFT L2-2 and L2-3 tests

  20. Microgravity experiments on a granular gas of elongated grains

    Science.gov (United States)

    Harth, K.; Trittel, T.; Kornek, U.; Höme, S.; Will, K.; Strachauer, U.; Stannarius, R.

    2013-06-01

    Granular gases represent well-suited systems to investigate statistical granular dynamics. The literature comprises numerous investigations of ensembles of spherical or irregularly shaped grains. Mainly computer models, analytical theories and experiments restricted to two dimensions were reported. In three-dimensions, the gaseous state can only be maintained by strong external excitation, e. g. vibrations or electro-magnetic fields, or in microgravity. A steady state, where the dynamics of a weakly disturbed granular gas are governed by particle-particle collisions, is hard to realize with spherical grains due to clustering. We present the first study of a granular gas of elongated cylinders in three dimensions. The mean free path is considerably reduced with respect to spheres at comparable filling fractions. The particles can be tracked in 3D over a sequence of frames. In a homogeneous steady state, we find non-Gaussian velocity distributions and a lack of equipartition of kinetic energy. We discuss the relations between energy input and vibrating plate accelerations. At the request of the authors and the Proceedings Editors, the PDF file of this article has been updated to amend some references present in the PDF file submitted to AIP Publishing. The references affected are listed here:[1] (c) K. Nichol and K. E. Daniels, Phys. Rev. Lett. 108, 018001 (2012); [11] (e) P. G. de Gennes and J. Prost, The Physics of Liquid Crystals, Clarendon Press, Oxford (1993); [17] (b) K. Harth, et al., Phys. Rev. Lett. 110, 144102 (2013).A LaTeX processing error resulted in changes to the authors reference formatting, which was not detected prior to publication. Due apologies are given to the authors for this oversight. The updated article PDF was published on 12 August 2013.

  1. Comparison of FISGAS swelling and gas release predictions with experiment

    International Nuclear Information System (INIS)

    Ostensen, R.W.

    1979-01-01

    FISGAS calculations were compared to fuel swelling data from the FD1 tests and to gas release data from the FGR39 test. Late swelling and gas release predictions are satisfactory if vacancy depletion effects are added to the code. However, early swelling predictions are not satisfactory, and early gas release predictions are very poor. Explanation of these discrepancies is speculative

  2. Velocity limitations in coaxial plasma gun experiments with gas mixtures

    International Nuclear Information System (INIS)

    Axnaes, I.

    1976-04-01

    The velocity limitations found in many crossed field plasma experiments with neutral gas present are studied for binary mixtures of H 2 , He, N 2 O 2 , Ne and Ar. The apparatus used is a coaxial plasma gun with an azimuthal magnetic bias field. The discharge parameters are chosen so that the plasma is weakly ionized. In some of the mixtures it is found that one of the components tends to dominate in the sense that only a small amount (regarding volume) of that component is needed for the discharge to adopt a limiting velocity close to that for the pure component. Thus in a mixture between a heavy and a light component having nearly equal ionization potentials the heavy component dominates. Also if there is a considerable difference in ionization potential between the components, the component with the lowest ionization potential tends to dominate. (author)

  3. Intercooler flow path for gas turbines: CFD design and experiments

    Energy Technology Data Exchange (ETDEWEB)

    Agrawal, A.K.; Gollahalli, S.R.; Carter, F.L. [Univ. of Oklahoma, Norman, OK (United States)] [and others

    1995-10-01

    The Advanced Turbine Systems (ATS) program was created by the U.S. Department of Energy to develop ultra-high efficiency, environmentally superior, and cost competitive gas turbine systems for generating electricity. Intercooling or cooling of air between compressor stages is a feature under consideration in advanced cycles for the ATS. Intercooling entails cooling of air between the low pressure (LP) and high pressure (BP) compressor sections of the gas turbine. Lower air temperature entering the HP compressor decreases the air volume flow rate and hence, the compression work. Intercooling also lowers temperature at the HP discharge, thus allowing for more effective use of cooling air in the hot gas flow path. The thermodynamic analyses of gas turbine cycles with modifications such as intercooling, recuperating, and reheating have shown that intercooling is important to achieving high efficiency gas turbines. The gas turbine industry has considerable interest in adopting intercooling to advanced gas turbines of different capacities. This observation is reinforced by the US Navys Intercooled-Recuperative (ICR) gas turbine development program to power the surface ships. In an intercooler system, the air exiting the LP compressor must be decelerated to provide the necessary residence time in the heat exchanger. The cooler air must subsequently be accelerated towards the inlet of the HP compressor. The circumferential flow nonuniformities inevitably introduced by the heat exchanger, if not isolated, could lead to rotating stall in the compressors, and reduce the overall system performance and efficiency. Also, the pressure losses in the intercooler flow path adversely affect the system efficiency and hence, must be minimized. Thus, implementing intercooling requires fluid dynamically efficient flow path with minimum flow nonuniformities and consequent pressure losses.

  4. Ultrasonic experiment on hydrate formation of a synthesis gas

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Shicai; Fan, Shuanshi; Liang, Deqing; Zhang, Junshe; Feng, Ziping

    2005-07-01

    The effect of ultrasonic on the induction time and formation rate of natural gas hydrates was investigated in a stainless steel cell in this study. The results show that the induction time with ultrasonic was about 1/6 of that without ultrasonic and only about 1/10 if rehydration after decomposition in water-gas system. In sodium dodecyl sulfate (SDS) solution-gas system, the critical micellar concentration (CMC) was not identified with ultrasonic. The formation rate and storage capacity of hydrate increased with increasing SDS concentration at a range of 0 to 800ppm. However, the increase was insignificant as the SDS concentration increased from 600 to 800ppm, (Author)

  5. Project Report of Virtual Experiments in Marine Bioacoustics: Model Validation

    Science.gov (United States)

    2010-08-01

    given initial velocity which starts the tissue " blobs " in opposing directions so that they collide and thereby produce an acoustic pressure wave. The...Atlantic bottlenose dolphin (Tursiops truncatus) in open waters. In: R. G. BUSNEL AND J. F. FISH (eds.). Animal Sonar Systems, pp. 251-282. Plenum

  6. Neutron radiography experiments for verification of soluble boron mixing and transport modeling under natural circulation conditions

    International Nuclear Information System (INIS)

    Morlang, M.M.; Feltus, M.A.

    1996-01-01

    The use of neutron radiography for visualization of fluid flow through flow visualization modules has been very successful. Current experiments at the Penn State Breazeale Reactor serve to verify the mixing and transport of soluble boron under natural flow conditions as would be experienced in a pressurized water reactor. Different flow geometries have been modeled including holes, slots, and baffles. Flow modules are constructed of aluminum box material 1 1/2 inches by 4 inches in varying lengths. An experimental flow system was built which pumps fluid to a head tank and natural circulation flow occurs from the head tank through the flow visualization module to be radio-graphed. The entire flow system is mounted on a portable assembly to allow placement of the flow visualization module in front of the neutron beam port. A neutron-transparent fluor-inert fluid is used to simulate water at different densities. Boron is modeled by gadolinium oxide powder as a tracer element, which is placed in a mixing assembly and injected into the system a remotely operated electric valve, once the reactor is at power. The entire sequence is recorded on real-time video. Still photographs are made frame-by-frame from the video tape. Computers are used to digitally enhance the video and still photographs. The data obtained from the enhancement will be used for verification of simple geometry predictions using the TRAC and RELAP thermal-hydraulic codes. A detailed model of a reactor vessel inlet plenum, downcomer region, flow distribution area and core inlet is being constructed to model the APGOO plenum. Successive radiography experiments of each section of the model under identical conditions will provide a complete vessel / core model for comparison with the thermal-hydraulic codes

  7. Neutron radiography experiments for verification of soluble boron mixing and transport modeling under natural circulation conditions

    International Nuclear Information System (INIS)

    Feltus, M.A.; Morlang, G.M.

    1996-01-01

    The use of neutron radiography for visualization of fluid flow through flow visualization modules has been very successful. Current experiments at the Penn State Breazeale Reactor serve to verify the mixing and transport of soluble boron under natural flow conditions as would be experienced in a pressurized water reactor. Different flow geometries have been modeled including holes, slots, and baffles. Flow modules are constructed of aluminum box material 1 1/2 inches by 4 inches in varying lengths. An experimental flow system was built which pumps fluid to a head tank and natural circulation flow occurs from the head tank through the flow visualization module to be radiographed. The entire flow system is mounted on a portable assembly to allow placement of the flow visualization module in front of the neutron beam port. A neutron-transparent fluorinert fluid is used to simulate water at different densities. Boron is modeled by gadolinium oxide powder as a tracer element, which is placed in a mixing assembly and injected into the system by remote operated electric valve, once the reactor is at power. The entire sequence is recorded on real-time video. Still photographs are made frame-by-frame from the video tape. Computers are used to digitally enhance the video and still photographs. The data obtained from the enhancement will be used for verification of simple geometry predictions using the TRAC and RELAP thermal-hydraulic codes. A detailed model of a reactor vessel inlet plenum, downcomer region, flow distribution area and core inlet is being constructed to model the AP600 plenum. Successive radiography experiments of each section of the model under identical conditions will provide a complete vessel/core model for comparison with the thermal-hydraulic codes

  8. Test Specifications and the Design of the Wire Wrapped 37-Pin Fuel Assembly for Hydrodynamic Experiments

    International Nuclear Information System (INIS)

    Chang, S. K.; Euh, D. J.; Bae, H.; Lee, H. Y.; Choi, S. R.

    2013-01-01

    Most influencing parameters on uncertainties and sensitivities of the CFD analyses are the friction coefficient and the mixing coefficient. The friction coefficient is related to the flow distribution in reactor sub-channels. The mixing coefficient is defined with the cross flow between neighboring sub-channels. The eventual purpose of the thermal hydraulic design considering these parameters is to guarantee the fuel cladding integrity as the design limit parameter. At the moment, the experimental program is being undertaken to quantify these friction and mixing parameters which characterize the flow distribution in sub-channels, and the wire wrapped 37-pin rod assembly and its hexagonal test rig have been designed and fabricated. The quantified thermal hydraulic experimental data from this program are utilized primarily to estimate the accuracy of the safety analysis codes and their thermal hydraulic model. A wire wrapped 37 pin fuel assembly has been designed for the measurements of the flow distribution, where the measurements are utilized to quantify the friction coefficient and the mixing coefficient. The test rig of the wire wrapped 37 pin fuel assembly has been fabricated considering the geometric and flow dynamic similarities. It comprises four components i. e., the upper plenum, the fuel housing, the lower plenum, and the wire wrapped 37 pin fuel assembly. At further works, the quantified friction and mixing coefficients through the experiments are going to be utilized for insuring the reliability of the CFD analysis results

  9. Testing and analyses of a high temperature thermal barrier for gas-cooled reactors

    International Nuclear Information System (INIS)

    Black, W.E.; Betts, W.S.; Felten, P.

    1979-01-01

    A full size, multi-panel section of a thermal barrier system was fabricated from a nickel-base superalloy and a combination of fibrous blanket insulation materials for specific application in a steam cycle gas-cooled nuclear reactor. The 2.4 m square array was representative of the sidewall of the lower core outlet plenum and included coverplates, attachments, seals, and a simulated water-cooled liner. Testing was conducted in a reactor grade, helium-filled chamber at 816 0 C for 100 hours, which established a normal (baseline) condition; 982 0 C for 10 hours, which satisfied an emergency condition; 1093 0 C for 1 hour, which simulated a faulted condition; and 1260 0 C, which was a non-design condition test to demonstrate the temperature overshoot capability of the system. Post-test examination indicated: (1) an acceptable performance by the anti-friction chromium carbide (Cr 3 C 2 ) coating; (2) no significant galling between non-coated surfaces; (3) no distortion of attachment fixtures; (4) predictable coverplate deflection during the design conditions testing (normal, emergency, and faulted); and (5) considerable plastic deformation resulting from the near-incipient melting temperature. (orig.)

  10. Modified laminar flow biological safety cabinet.

    Science.gov (United States)

    McGarrity, G J; Coriell, L L

    1974-10-01

    Tests are reported on a modified laminar flow biological safety cabinet in which the return air plenum that conducts air from the work area to the high efficiency particulate air filters is under negative pressure. Freon gas released inside the cabinet could not be detected outside by a freon gas detection method capable of detecting 10(-6) cc/s. When T3 bacteriophage was aerosolized 5 cm outside the front opening in 11 tests, no phage could be detected inside the cabinet with the motor-filter unit in operation. An average of 2.8 x 10(5) plaque-forming units (PFU)/ft(3) (ca. 0.028 m(3)) were detected with the motor-filter unit not in operation, a penetration of 0.0%. Aerosolization 5 cm inside the cabinet yielded an average of 10 PFU/ft(3) outside the cabinet with the motor-filter unit in operation and an average of 4.1 x 10(5) PFU/ft(3) with the motor-filter unit not in operation, a penetration of 0.002%. These values are the same order of effectiveness as the positive-pressure laminar flow biological safety cabinets previously tested. The advantages of the negative-pressure return plenum design include: (i) assurance that if cracks or leaks develop in the plenum it will not lead to discharge of contaminated air into the laboratory; and (ii) the price is lower due to reduced manufacturing costs.

  11. Dish/stirling hybrid-receiver

    Science.gov (United States)

    Mehos, Mark S.; Anselmo, Kenneth M.; Moreno, James B.; Andraka, Charles E.; Rawlinson, K. Scott; Corey, John; Bohn, Mark S.

    2002-01-01

    A hybrid high-temperature solar receiver is provided which comprises a solar heat-pipe-receiver including a front dome having a solar absorber surface for receiving concentrated solar energy, a heat pipe wick, a rear dome, a sidewall joining the front and the rear dome, and a vapor and a return liquid tube connecting to an engine, and a fossil fuel fired combustion system in radial integration with the sidewall for simultaneous operation with the solar heat pipe receiver, the combustion system comprising an air and fuel pre-mixer, an outer cooling jacket for tangentially introducing and cooling the mixture, a recuperator for preheating the mixture, a burner plenum having an inner and an outer wall, a porous cylindrical metal matrix burner firing radially inward facing a sodium vapor sink, the mixture ignited downstream of the matrix forming combustion products, an exhaust plenum, a fossil-fuel heat-input surface having an outer surface covered with a pin-fin array, the combustion products flowing through the array to give up additional heat to the receiver, and an inner surface covered with an extension of the heat-pipe wick, a pin-fin shroud sealed to the burner and exhaust plenums, an end seal, a flue-gas diversion tube and a flue-gas valve for use at off-design conditions to limit the temperature of the pre-heated air and fuel mixture, preventing pre-ignition.

  12. Investigation of alpha experiment by severe accident analysis code SAMPSON

    International Nuclear Information System (INIS)

    Baglietto, Emilio; Ninokata, Hisashi; Naitoh, Masanori

    2006-01-01

    The severe accident analysis code SAMPSON is adopted in this work to evaluate its capability of reproducing the complex gap cooling phenomenon. The ALPHA experiment is adopted for validation, where molten aluminum oxide (Al 2 O 3 ) produced by a thermite reaction is poured into a water filled hemispherical vessel at the ambient pressure of approximately 1.3 MPa. The spreading and cooling of the debris that has relocated into the pressure vessel lower plenum are simulated, including the analysis of the RPV failure. The model included in the core to mimic the water penetration inside the gap is evaluated and improvements are proposed. The importance of the introduction of some mechanistic approach to describe the gap formation and evolution is underlined, where the results show its necessity in order to correctly reproduce the experimental trends. (author)

  13. Fluid Phase Separation (FPS) experiment for flight on a space shuttle Get Away Special (GAS) canister

    Science.gov (United States)

    Peters, Bruce; Wingo, Dennis; Bower, Mark; Amborski, Robert; Blount, Laura; Daniel, Alan; Hagood, Bob; Handley, James; Hediger, Donald; Jimmerson, Lisa

    1990-01-01

    The separation of fluid phases in microgravity environments is of importance to environmental control and life support systems (ECLSS) and materials processing in space. A successful fluid phase separation experiment will demonstrate a proof of concept for the separation technique and add to the knowledge base of material behavior. The phase separation experiment will contain a premixed fluid which will be exposed to a microgravity environment. After the phase separation of the compound has occurred, small samples of each of the species will be taken for analysis on the Earth. By correlating the time of separation and the temperature history of the fluid, it will be possible to characterize the process. The experiment has been integrated into space available on a manifested Get Away Special (GAS) experiment, CONCAP 2, part of the Consortium for Materials Complex Autonomous Payload (CAP) Program, scheduled for STS-42. The design and the production of a fluid phase separation experiment for rapid implementation at low cost is presented.

  14. Pennsylvania's technologically enhanced, naturally occurring radioactive material experiences and studies of the oil and gas industry.

    Science.gov (United States)

    Allard, David J

    2015-02-01

    This presentation provides an overview of the Commonwealth of Pennsylvania's experiences and ongoing studies related to technologically enhanced, naturally occurring radioactive material (TENORM) in the oil and gas industry. It has been known for many years that Pennsylvania's geology is unique, with several areas having relatively high levels of natural uranium and thorium. In the 1950s, a few areas of the state were evaluated for commercial uranium production. In the late 1970s, scoping studies of radon in homes prompted the Pennsylvania Department of Environmental Protection (DEP) Bureau of Radiation Protection (BRP) to begin planning for a larger state-wide radon study. The BRP and Oil and Gas Bureau also performed a TENORM study of produced water in the early 1990s for a number of conventional oil and gas wells. More recently, BRP and the Bureau of Solid Waste developed radiation monitoring regulations for all Pennsylvania solid waste disposal facilities. These were implemented in 2001, prompting another evaluation of oil and gas operations and sludge generated from the treatment of conventionally produced water and brine but mainly focused on the disposal of TENORM solid waste in the state's Resource Conservation and Recovery Act Subtitle D landfills. However, since 2008, the increase in volumes of gas well wastewater and levels of Ra observed in the unconventional shale gas well flow-back fracking water has compelled DEP to fully re-examine these oil and gas operations. Specifically, with BRP in the lead, a new TENORM study of oil and gas operations and related wastewater treatment operations has been initiated (), supported by an American National Standards Institute standard on TENORM () and a U.S. Government Accountability Office report on shale resource development and risks (). This study began in early 2013 and will examine the potential public and worker radiation exposure and environmental impact as well as re-evaluate TENORM waste disposal. This

  15. Hydrogen gas detector

    International Nuclear Information System (INIS)

    Bohl, T.L.

    1982-01-01

    A differential thermocouple hydrogen gas detector has one thermocouple junction coated with an activated palladium or palladium-silver alloy catalytic material to allow heated hydrogen gas to react with the catalyst and raise the temperature of that junction. The other juction is covered with inert glass or epoxy resin, and does not experience a rise in temperature in the presence of hydrogen gas. A coil heater may be mounted around the thermocouple junctions to heat the hydrogen, or the gas may be passed through a heated block prior to exposing it to the thermocouples

  16. Measurements of electron density and temperature profiles in a gas blanket experiment

    International Nuclear Information System (INIS)

    Kuthy, A.

    1979-02-01

    Radial profiles of electron density, temperature and H sub(β) intensity are presented for the rotating plasma device F-1. The hydrogen filling pressure, the average magnetic field strength at the midplane, and the power input to the discharge have been varied in the ranges 10-100 mTorr, 0.25-0.5 Tesla, and 0.1 to 1.5 MW, respectively. These experiments have been performed with the main purpose of studying the gas blanket (cold-mantle) state of the plasma. It is shown, that a simple spectroscopic method can be used to derive the radial distribution of the electron temperature in such plasmas. The observed peak temperatures and densities are in agreement with earlier theoretical estimates. (author)

  17. Correlation of Steam Generator Mixing Parameters for Severe Accident Hot-Leg Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Liao, Yehong; Guentay, Salih [Paul Scherrer Institut, Villigen PSI, CH-5232 (Switzerland)

    2008-07-01

    Steam generator inlet plenum mixing phenomenon with hot-leg counter-current natural circulation during a PWR station blackout severe accident is one of the important processes governing which component will fail first as a result of thermal challenge from the circulating gas with high temperature and pressure. Since steam generator tube failure represents bypass release of fission product from the reactor to environment, study of inlet plenum mixing parameters is important to risk analysis. Probability distribution functions of individual mixing parameter should be obtained from experiments or calculated by analysis. In order to perform sensitivity studies of the synergetic effects of all mixing parameters on the severe accident-induced steam generator tube failure, the distribution and correlation of these mixing parameters must be known to remove undue conservatism in thermal-hydraulic calculations. This paper discusses physical laws governing three mixing parameters in a steady state and setups the correlation among these mixing parameters. The correlation is then applied to obtain the distribution of one of the mixing parameters that has not been given in the previous CFD analysis. Using the distributions and considering the inter-dependence of the three mixing parameters, three sensitivity cases enveloping the mixing parameter uncertainties are recommended for the plant analysis. (authors)

  18. Physics of gas breakdown for ion beam transport in gas

    International Nuclear Information System (INIS)

    Olson, C.L.; Poukey, J.W.; Hinshelwood, D.D.; Rose, D.V.; Hubbard, R.F.; Lampe, M.; Neri, J.M.; Ottinger, P.F.; Slinker, S.P.; Stephanakis, S.J.; Young, F.C.; Welch, D.R.

    1993-01-01

    Detailed analysis, experiments, and computer simulations are producing a new understanding of gas breakdown during intense ion beam transport in neutral gas. Charge neutralization of beam micro clumps is shown to limit the net clump potentials to a non-zero value π min , which can lead to divergence growth and axial energy spreading. At pressures approx-gt 1 Torr, plasma shielding should substantially reduce this effect Current neutralization has been studied in experiments on the GAMBLE II accelerator. The importance of fast electrons (knockons and runaways) has been established in IPROP simulations, which are in agreement with the experiments. For light ion fusion parameters with pressures approx-gt 1 Torr, very small net current fractions (much-lt 1%) appear feasible, permitting ballistic transport in gas. Self-pinched requires higher net current fractions (≥ 2%) and preliminary IPROP code results indicate that this appears achievable for small-radius intense beams in lower pressure gases (approx-gt Torr). Several self-pinched transport concepts look promising. The importance of these results for both light ion fusion and heavy ion fusion is discussed

  19. 40 CFR 63.11511 - What definitions apply to this subpart?

    Science.gov (United States)

    2010-07-01

    ... detonation gun spraying. Water curtain means a type of control device that draws the exhaust stream through a... emissions points and then convey the captured gas stream to a control device, as part of a complete control..., plenums, and fans. Cartridge filter means a type of control device that uses perforated metal cartridges...

  20. Planning of in-situ experiment for understanding of gas migration behaviour in sedimentary rock. (1) Setting of gas injection procedure

    International Nuclear Information System (INIS)

    Tanai, Kenji; Fujita, Tomoo; Noda, Masaru; Yamamoto, Shuichi; Shimura, Tomoyuki; Sato, Shin

    2013-01-01

    Japan Atomic Energy Agency has been planning in-situ gas migration test in Horonobe URL, Hokkaido. This paper discusses the optimum gas injection procedure for the test to understand gas migration behaviour in surrounded rock. The stepwise constant gas injection was selected, taking into account domestic and overseas gas related research results. Hydro-mechanical-gas coupling analysis which is able to consider the dissolved methane in Horonobe groundwater was applied to evaluate the gas behaviour. The results have indicated no significant mechanical damages to the rock and have supported the appropriateness of selected gas injection procedure for the test. (author)

  1. Investigation on cold fusion phenomena using gas-metal loading experiments

    International Nuclear Information System (INIS)

    Lanza, F.; Bertolini, G.; Vocino, V.; Parnisari, E.; Ronsecco, C.

    1992-01-01

    Previous experiments have shown that tritium is produced in deuterated titanium. The data obtained are highly scattered and non reproducible. In order to try to define better the phenomenon a series of tests have been performed using various metals and alloys and different deuterating conditions. Sheets and shavings of titanium, zirconium, hafnium, tantalum, zircaloy 2 and Ti-Zr 5O% alloy have been tested. The tritium production is obtained as a difference of the tritium content in the deuterated metal and the initial content of tritium in the deuterium gas. The amount of tritium produced is low and reproducibility is rather poor. A statistical analysis shows that significant differences are obtained varying the type of metal used. In general the tritium production increases with the atomic number of the metal. Moreover significantly higher productions of tritium have been obtained using materials of technical purity as tantalum, zircaloy 2 and Ti-Zr alloy

  2. HRL Aespoe - two-phase flow experiment - gas and water flow in fractured crystalline rock

    International Nuclear Information System (INIS)

    Kull, H.; Liedtke, L.

    1998-01-01

    (The full text of the contribution follows:) Gas generated from radioactive waste may influence the hydraulic and mechanical properties of the man-made barriers and the immediate surroundings of the repository. Prediction of alteration in fractured crystalline rock is difficult. There is a lack of experimental data, and calibrated models are not yet available. Because of the general importance of this matter the German Federal Ministry for Education, Science, Research and Technology decided to conduct a two-phase flow study at HRL Aespoe within the scope of the co-operation agreement with SKB. Within the presentation an overview of field experiments and modelling studies scheduled until end of '99 are given. Conceptual models for one- and two-phase flow, methodologies and with respect to numerical calculations necessary parameter set-ups are discussed. Common objective of in-situ experiments is to calibrate flow models to improve the reliability of predictions for gas migration through fractured rock mass. Hence, in a defined dipole flow field in niche 2/715 at HRL Aespoe effective hydraulic parameters are evaluated. Numerical modelling of non-isothermal, two-phase, two-component processes is feasible only for two-dimensional representation of a porous medium. To overcome this restriction a computer program will be developed to model three-dimensional, fractured, porous media. Rational aspects of two-phase flow studies are for the designing of geotechnical barriers and for the long-term safety analysis of potential radionuclide transport in a future repository required for the licensing process

  3. Simulations of the Viking Gas Exchange Experiment using palagonite and Fe-rich montmorillonite as terrestrial analogs: implications for the surface composition of Mars.

    Science.gov (United States)

    Quinn, R; Orenberg, J

    1993-10-01

    Simulations of the Gas Exchange Experiment (GEX), one of the Viking Lander Biology Experiments, were run using palagonite and Fe-rich montmorillonite as terrestrial analogs of the Martian soil. These terrestrial analogs were exposed to a nutrient solution of the same composition as that of the Viking Landers under humid (no contact with nutrient) and wet (intimate contact) conditions. The headspace gases in the GEX sample cell were sampled and then analyzed by gas chromatography under both humid and wet conditions. Five gases were monitored: CO2, N2, O2, Ar, and Kr. It was determined that in order to simulate the CO2 gas changes of the Viking GEX experiment, the mixture of soil analog mineral plus nutrient medium must be slightly (pH = 7.4) to moderately basic (pH = 8.7). This conclusion suggests constraints upon the composition of terrestrial analogs to the Mars soil; acidic components may be present, but the overall mixture must be basic in order to simulate the Viking GEX results.

  4. The spray characteristic of gas-liquid coaxial swirl injector by experiment

    OpenAIRE

    Chen Chen; Zhihui Yan; Yang Yang; Hongli Gao; Shunhua Yang; Lei Zhang

    2017-01-01

    Using the laser phase Doppler particle analyzer (PDPA), the spray characteristics of gas-liquid coaxial swirl injector were studied. The Sauter mean diameter (SMD), axial velocity and size data rate were measured under different gas injecting pressure drop and liquid injecting pressure drop. Comparing to a single liquid injection, SMD with gas presence is obviously improved. So the gas presence has a significant effect on the atomization of the swirl injector. What’s more, the atomization eff...

  5. Axial gas transport and loss of pressure after ballooning rupture of high burn-up fuel rods subjected to LOCA conditions

    International Nuclear Information System (INIS)

    Wiesenack, Wolfgang; Oberlaender, Barbara; Kekkonen, Laura

    2008-01-01

    The OECD Halden Reactor Project has implemented integral in-pile tests on issues related to fuel behaviour under LOCA conditions. In this test series, the interaction of bonded fuel and cladding, the behaviour of fragmented fuel around the ballooning area, and the axial gas communication in high burn-up rods as affected by gap closure and fuel-clad bonding are of major interest for the investigations. In the Halden reactor tests, the decay heat is simulated by a low level of nuclear heating, in contrast to the heating conditions implemented in hot laboratory set-ups, and the thermal expansion of fuel and cladding relative to each other is more similar to the real event. The paper deals with observations regarding the loss of rod pressure following the rupture of the cladding. In the majority of the tests conducted so far, the rod pressure dropped practically instantaneously as a consequence of ballooning rupture, while one test showed a remarkably slow pressure loss. The slow loss of pressure in this test was analysed, showing that the 'hydraulic diameter' of the rod over an un-distended upper part was about 30 - 35 μm which is typical of high burn-up fuel at hot-standby conditions. The 'plug' of fuel restricts the gas flow from the plenum through the fuel column and thus limits the availability of high pressure gas for driving the ballooning. This observation is relevant for the analysis of the behaviour of a full length fuel rod under LOCA conditions since restricted gas flow may influence bundle blockage and the number of failures. (authors)

  6. Design/build/mockup of the Waste Isolation Pilot Plant gas generation experiment glovebox

    International Nuclear Information System (INIS)

    Rosenberg, K.E.; Benjamin, W.W.; Knight, C.J.; Michelbacher, J.A.

    1996-01-01

    A glovebox was designed, fabricated, and mocked-up for the WIPP Gas Generation Experiments (GGE) being conducted at ANL-W. GGE will determine the gas generation rates from materials in contact handled transuranic waste at likely long term repository temperature and pressure conditions. Since the customer's schedule did not permit time for performing R ampersand D of the support systems, designing the glovebox, and fabricating the glovebox in a serial fashion, a parallel approach was undertaken. As R ampersand D of the sampling system and other support systems was initiated, a specification was written concurrently for contracting a manufacturer to design and build the glovebox and support equipment. The contractor understood that the R ampersand D being performed at ANL-W would add additional functional requirements to the glovebox design. Initially, the contractor had sufficient information to design the glovebox shell. Once the shell design was approved, ANL-W built a full scale mockup of the shell out of plywood and metal framing; support systems were mocked up and resultant information was forwarded to the glovebox contractor to incorporate into the design. This approach resulted in a glovebox being delivered to ANL-W on schedule and within budget

  7. Experiments on state selection and Penning ionisation with fast metastable rare gas atoms

    International Nuclear Information System (INIS)

    Kroon, J.P.C.

    1985-01-01

    This thesis describes experiments with metastable He/Ne atoms. The experiments are performed in a crossed beam machine. Two different sources are used for the production of metastable atoms: a source for the production of metastable atoms in the thermal energy range and a hollow cathode arc for the production of metastable atoms in the superthermal energy range (1-7 eV). The progress made in the use of the hollow cathode arc is described as well as the experimental set-up. The rare gas energy-level diagram is characterized by two metastable levels. By optical pumping it is possible to select a single metastable level, both for He and Ne. For the case of He this is done by a recently built He quenchlamp which selectively quenches the metastable 2 1 S level population. In the thermal energy range the quenching is complete; in the superthermal energy range the 2 1 S level population is only partly quenched. For the optical pumping of Ne* atoms a cw dye laser is used. New experiments have been started on the measurement, in a crossed beam machine, of the fluorescence caused by inelastic collisions where metastable atoms are involved. The He* + Ne system is used as a pilot study for these experiments. The He-Ne laser is based on this collision system. (Auth.)

  8. Gas-puff Z-pinch experiment on the LIMAY-I

    International Nuclear Information System (INIS)

    Takasugi, K.; Miyamoto, T.; Akiyama, H.; Shimomura, N.; Sato, M.; Tazima, T.

    1989-01-01

    A gas-puff z-pinch plasma has been produced on the pulsed power generator LIMAY-I at IPP Nagoya University. The stored energy of the generator is 13 kJ, and it generates 600 kV-70 ns-3 Ω power pulse. Ar or He gas is puffed from a hollow nozzle with 18 mm diameter, and a z-pinch plasma is produced by a discharge between 3 mm gap electrodes

  9. Experiment data report for Semiscale Mod-1 Test S-05-1 (alternate ECC injection test)

    International Nuclear Information System (INIS)

    Feldman, E.M.; Patton, M.L. Jr.; Sackett, K.E.

    1977-02-01

    Recorded test data are presented for Test S-05-1 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-1 was conducted from initial conditions of 2263 psia and 544 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the vessel lower plenum to simulate emergency core coolant injection in a PWR, with the flow rate based on system volume scaling

  10. Flow characteristics of centrifugal gas-liquid separator. Investigation with air-water two-phase flow experiment

    International Nuclear Information System (INIS)

    Yoneda, Kimitoshi; Inada, Fumio

    2004-01-01

    Air-water two-phase flow experiment was conducted to examine the basic flow characteristics of a centrifugal gas-liquid separator. Vertical transparent test section, which is 4 m in height, was used to imitate the scale of a BWR separator. Flow rate conditions of gas and liquid were fixed at 0.1 m 3 /s and 0.033 m 3 /s, respectively. Radial distributions of two-phase flow characteristics, such as void fraction, gas velocity and bubble chord length, were measured by traversing dual optical void probes in the test section, horizontally. The flow in the standpipe reached to quasi-developed state within the height-to-diameter aspect ratio H/D=10, which in turn can mean the maximum value for an ideal height design of a standpipe. The liquid film in the barrel showed a maximum thickness at 0.5 to 1 m in height from the swirler exit, which was a common result for three different standpipe length conditions, qualitatively and quantitatively. The empirical database obtained in this study would contribute practically to the validation of numerical analyses for an actual separator in a plant, and would also be academically useful for further investigations of two-phase flow in large-diameter pipes. (author)

  11. Dynamic structural response of reactor-core subassemblies (hexcans) due to accident overpressurization

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kulak, R.F.

    1993-01-01

    This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall. (author)

  12. Dynamic structural response of reactor-core subassemblies (hexcans) due to accident overpressurization

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kulak, R.F.

    1993-01-01

    This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall

  13. Structural response of reactor-core hexcan subassemblies subjected to dynamic overpressurization under accident conditions

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kulak, R.F.

    1993-01-01

    This paper presents a two-dimensional structural analysis for the evaluation of a single core subassembly due to internal overpressure associated with possible failure of fuel pins having high fission gas plenum pressure. Structural models are developed for the subassemblies and their surroundings with emphasis on the critical physical aspects of the problem. With these models the strains, deformations and the extent of permanent damage (plastic strain) to the subassemblies can be assessed. The nonlinear structural analyses was performed with a finite element program called STRAW (Structural Transient Response of Assembly Wrappers). This finite element program is applicable to nonlinear large displacement problems. The results of this study indicate that the permanent deformation (damage) is strongly influenced by the rise time (time to reach peak pressure) of the pressure pulse and the pressure in the fuel pin. The rise time is influenced by the opening time of the flow path for release of gas from the fuel pin plenum. Several examples are illustrated with various rise times and pressure magnitudes and the resulting permanent deformation of the hexcan wall

  14. Current gas storage R and D programmes at Gas Research Institute

    International Nuclear Information System (INIS)

    Shikari, Y.A.

    1990-01-01

    The Gas Research Institute (GRI) is currently involved in the development of concepts aimed at an enhancement of natural gas service to the consumer. In order to maintain the attractiveness of the gas options to industrial consumers and to reinforce the ''value-in-use'' of natural gas to residential as well as commercial customers, it is essential to develop efficient, economical, and safe means of reducing the ''cost-of-service'', including that of natural gas storage in underground formations. Specifically, research and development (R and D) is needed to explore ways to better utilize existing storage fields and also to develop new storage facilities at minimum cost. GRI is currently sponsoring research projects aimed at controlling gas migration in underground gas storage reservoirs, reducing base (or cushion) gas requirements, understanding the gas-gas phase mixing behaviour via laboratory experiments and reservoir models, developing cost-effective gas separation processes using membranes, and optimizing the operation and maintenance (O and M) costs of underground gas storage operations. This paper provides an overview of the GRI's Gas Storage R and D Programme and highlights key results achieved to date for selected research projects. (author). 16 refs, 6 figs, 3 tabs

  15. Operating experience of gas bearing helium circulators in HTGR development facility

    International Nuclear Information System (INIS)

    Shimomura, H.; Kawaji, S.; Fujisaki, K.; Ihizuka, T.

    1988-01-01

    The large scale helium gas test facility (HENDEL) has been constructed and operated since March 1982 at the Japan Atomic Energy Research Institute to develop HTGR components. The five electric driven gas circulators with dynamic gas bearings are used to circulate the helium gas of 4MPa and 400 deg. C in loops for their compactness, gas tightness, easy maintenance and free from gas contamination. All of these circulators are variable speed types of 3,000 to 12,000 rpm and have the same gas bearings and electric motors. The four machines among them are equipped with centrifugal impeller and one other machine has regenerative type, and the weight of both type rotors are nearly the same. After the troubles and repairing, both type of circulators were tested and the vibration characteristics were measured as preventing maintenance. From the test and measurements of the circulators, it was presumed that the first trouble on regenerative type was caused from excess unbalance force by falling off of a small pin from the rotating part and the second severe trouble on it was caused by the whipping in gas bearing. The static load on tilting pads indicated close relations to occurrence of the whirling through the measurements. It is recognized that fine balancing of the rotors and delicate clearance adjustment of the bearings are very important for the rotor stability and that the mechanism should be designed and machined so precise as to be adjustable. As the gas bearing would be damaged in an instantaneously short time, the monitoring technique for it should be so fast and predictive as to prevent serious damage. Through the tests, the vibration spectrum monitoring method seems to be predictive and useful for early detection of the shaft instability. It will be concluded that the gas bearing machine is an excellent system in its design philosophy, however, it also needs highly precise machining and delicate maintenance technique. 4 refs, 10 figs, 1 tab

  16. Dual-purpose power plants, experiences with exhaust gas purification plants

    International Nuclear Information System (INIS)

    Dietrich, R.

    1993-01-01

    From 1984 to 1988, the research and development project ''pollutant reduction for exhaust gases from heat production systems'' sponsored by the Federal Ministry of Research and Technology (BMFT) has been carried out by TUeV in Bavaria. This project was to show the state of exhaust gas technology for small and medium-sized plants (boilers and motoric heat generators). When publishing the final report, no positive balance could be given. Based on the results, the succession project ''Exhaust gas purification plants in field test'' (ARIF) has been started. This project has the following objectives: -Measuring technical investigation of the exhaust gas purification of motoric driven heat generator systems in field test. - Suitability of hand measuring devices for emissions for a discontinuous control of the exhaust gas purification plat by the operator. - Control of new methods regarding pollutant reduction for motoric and conventional heat generators. (orig.) [de

  17. Effects of gas composition in headspace and bicarbonate concentrations in media on gas and methane production, degradability, and rumen fermentation using in vitro gas production techniques.

    Science.gov (United States)

    Patra, Amlan Kumar; Yu, Zhongtang

    2013-07-01

    Headspace gas composition and bicarbonate concentrations in media can affect methane production and other characteristics of rumen fermentation in in vitro gas production systems, but these 2 important factors have not been evaluated systematically. In this study, these 2 factors were investigated with respect to gas and methane production, in vitro digestibility of feed substrate, and volatile fatty acid (VFA) profile using in vitro gas production techniques. Three headspace gas compositions (N2+ CO2+ H2 in the ratio of 90:5:5, CO2, and N2) with 2 substrate types (alfalfa hay only, and alfalfa hay and a concentrate mixture in a 50:50 ratio) in a 3×2 factorial design (experiment 1) and 3 headspace compositions (N2, N2 + CO2 in a 50:50 ratio, and CO2) with 3 bicarbonate concentrations (80, 100, and 120 mM) in a 3×3 factorial design (experiment 2) were evaluated. In experiment 1, total gas production (TGP) and net gas production (NGP) was the lowest for CO2, followed by N2, and then the gas mixture. Methane concentration in headspace gas after fermentation was greater for CO2 than for N2 and the gas mixture, whereas total methane production (TMP) and net methane production (NMP) were the greatest for CO2, followed by the gas mixture, and then N2. Headspace composition did not affect in vitro digestibility or the VFA profile, except molar percentages of propionate, which were greater for CO2 and N2 than for the gas mixture. Methane concentration in headspace gas, TGP, and NGP were affected by the interaction of headspace gas composition and substrate type. In experiment 2, increasing concentrations of CO2 in the headspace decreased TGP and NGP quadratically, but increased the concentrations of methane, NMP, and in vitro fiber digestibility linearly, and TMP quadratically. Fiber digestibility, TGP, and NGP increased linearly with increasing bicarbonate concentrations in the medium. Concentrations of methane and NMP were unaffected by bicarbonate concentration, but

  18. An assessment of gas impact on geological repository. Methodology and material property of gas migration analysis in engineered barrier system

    International Nuclear Information System (INIS)

    Yamamoto, Mikihiko; Mihara, Morihiro; Ooi, Takao

    2004-01-01

    Gas production in a geological repository has potential hazard, as overpressurisation and enhanced release of radionuclides. Amongst data needed for assessment of gas impact, gas migration properties of engineered barriers, focused on clayey and cementitious material, was evaluated in this report. Gas injection experiments of saturated bentonite sand mixture, mortar and cement paste were carried out. In the experiments, gas entry phenomenon and gas outflow rate were observed for these materials. Based on the experimental results, two-phase flow parameters were evaluated quantitatively. A conventional continuum two-phase flow model, which is only practically used multidimensional multi-phase flow model, was applied to fit the experimental results. The simulation results have been in good agreement with the gas entry time and the outflow flux of gas and water observed in the experiments. It was confirmed that application of the continuum two-phase flow model to gas migration in cementitious materials provides sufficient degree of accuracy for assessment of repository performance. But, for sand bentonite mixture, further extension of basic two-phase flow model is needed especially for effect of stress field. Furthermore, gas migration property of other barrier materials, including rocks, but long-term gas injection test, clarification of influence of chemicals environment and large-scale gas injection test is needed for multi-barrier assessment tool development and their verification. (author)

  19. Fully coupled fluid-structure interaction model of reed valves in a multi-cylinder reciprocating piston compressor

    Science.gov (United States)

    Xie, F.; Nieter, J.; Lifson, A.; Reba, R.; Sishtla, V.

    2017-08-01

    For years compressor researchers have tried to account for the fluid interaction effect of the working fluid on valve motion in displacement compressors. In recent years, the computing capacities and available CFD and FEA programs have allowed fully coupled interaction of fluids and moving structures to be modelled more comprehensively. This paper describes our experience and results from developing a model of a multi-cylinder reciprocating piston compressor with suction and discharge valve systems that are fully coupled with the pressure pulsation in the adjacent plenum. Valve dynamics are captured by the model that affects compressor performance. The results show that higher running speed causes more discharge valve delay on closing due to higher pressure pulsation in discharge plenum. The acoustic property of the discharge plenum as it relates to valve motion is studied by the developed cost-effective standalone model.

  20. Water Resource Impacts During Unconventional Shale Gas Development: The Pennsylvania Experience

    Science.gov (United States)

    Brantley, S. L.; Yoxtheimer, D.; Arjmand, S.; Grieve, P.; Vidic, R.; Abad, J. D.; Simon, C. A.; Pollak, J.

    2013-12-01

    The number of unconventional Marcellus shale wells in PA has increased from 8 in 2005 to more than 6000 today. This rapid development has been accompanied by environmental issues. We analyze publicly available data describing this Pennsylvania experience (data from www.shalenetwork.org and PA Department of Environmental Protection, i.e., PA DEP). After removing permitting and reporting violations, the average percent of wells/year with at least one notice of violation (NOV) from PA DEP is 35 %. Most violations are minor. An analysis of NOVs reported for wells drilled before 2013 revealed a rate of casing, cement, or well construction issues of 3.4%. Sixteen wells were given notices specifically related to migration of methane. A similarly low percent of wells were contaminated by brine components. Such contamination could derive from spills, subsurface migration of flowback water or shallow natural brines, or contamination by drill cuttings. Most cases of contamination of drinking water supplies with methane or brine components were reported in the previously glaciated part of the state. Before 2011, flowback and production water was often discharged legally into streams after minimal treatment, possibly increasing dissolved Br concentrations in some rivers. The rate of large spills or releases of gas-related industrial wastes in the state peaked in 2009 but little evidence of spills has been found in publicly available surface water chemistry data. The most likely indicators of spillage or subsurface release of flowback or production waters are the dissolved ions Na, Ca, and Cl. However, the data coverage for any given analyte is generally spatially and temporally sparse. Publicly available water quality data for before and after spills into Larrys Creek and Bobs Creek document the difficulties of detecting such events. An observation from the Pennsylvania experience is that the large number of people who have complained about their water supply (~1000 letters

  1. RELAP5/MOD 3.2 Analysis of the Loss of RHR System Experiment Scaled to NPP Krsko

    International Nuclear Information System (INIS)

    Bencik, V.; Bajs, T.; Prah, M.

    1998-01-01

    In the paper the RELAP5/MOD 3.2 analysis of the loss of Residual Heat Removal (RHR) system during midloop operation experiment performed at the Rig of Safety Assessment (ROSA)-IV/Large Scale Test Facility (LSTF) together with the analysis of the same test scenario scaled to NPP Krsko are presented. The experiment consisted in a loss of the RHR system at cold shutdown conditions along with a 5% cold leg break in the loop without pressurizer. The Safety Injection (SI) system was disable in the calculation. The aims of the work were to study the physical phenomena encountered under low power and low system pressure conditions while the upper part of the Reactor Coolant System (RCS) is filled with noncondensable. The impact of the bypass flow between upper plenum and downcomer inlet on transient responses was investigated. The transient was simulated for 6000 s. (author)

  2. Design, fabrication and testing of the gas analysis system for the tritium recovery experiment, TRIO-01

    International Nuclear Information System (INIS)

    Finn, P.A.; Reedy, G.T.; Homa, M.I.; Clemmer, R.G.; Pappas, G.; Slawecki, M.A.; Graczyk, D.G.; Bowers, D.L.; Clemmer, E.D.

    1983-01-01

    The tritium recovery experiment, TRIO-01, required a gas analysis system which detected the form of tritium, the amount of tritium (differential and integral), and the presence and amount of other radioactive species. The system had to handle all contingencies and function for months at a time unattended during weekend operation. The designed system, described herein, consisted of a train of components which could be grouped as desired to match tritium release behavior

  3. Studies of two stage gas turbine combustor for biomass powder. Part 1, Atmospheric cyclone gasification experiments with wood powder. Technical report

    Energy Technology Data Exchange (ETDEWEB)

    Degerman, Bengt; Hedin, Johan; Fredriksson, Christian; Kjellstroem, Bjoern; Salman, Hassan [Luleaa Univ. of Technology (Sweden). Dept. of Mechanical Engineering

    2000-10-01

    This report summarises the research and development work regarding development of a two stage gas turbine combustor for wood powder carried out at the Luleaa University of Technology from July 1993 to December 1996. The process being studied is based on cyclone gasification of the wood powder and combustion of the product gas in a suitably adapted gas turbine combustion chamber, without other gas cleaning than that obtained by the cyclone. A critical issue to be studied in the project is if the burned gases from such a cyclone gasifier lead to acceptably low deposition rates for K- and Na-compounds in a gas turbine with 850 deg C inlet temperature. The project strategy has been to study wood powder feeding and cyclone gasification first at atmospheric pressure, then run separate pressurised cyclone gasification tests for studies of the possibilities to achieve stable operation when the air flow is supplied by a separate compressor and finally to run integrated gasifier/gas turbine tests for studies of the deposition problem in practical operation. During the period covered by this report the atmospheric test facility has been designed, built and commissioned. It has been used mainly for studies of injector feeding of wood powder into a cyclone gasifier and for gasification experiments where in particular the fate of ash elements introduced with the wood powder has been studied. The results of these experiments have shown that steam injection of wood powder is possible with a steam consumption of about 0.3 kg steam/kg wood. The effects of injector geometry on the performance has also been studied. The gasification experiments show clearly that ash elements, including K and Na remain in the ash until very late in the thermal conversion process, also at gas temperatures exceeding 900 deg C. The separation of K with the cyclone bottom char has been 50 - 60% and the separation of Na about 80% with the cyclone geometry and the wood powder tested. The resulting load of K

  4. Focusing an antimatter beam with matter

    CERN Document Server

    CERN. Geneva

    2000-01-01

    An experiment at the Stanford Linear Accelerator Center has recently focused positron beams by means of a plasma lens. This is the first time this process has been observed. The process started with a positron beam from the SLAC PEP-II positron source. This was sent through a damping ring and then accelerated to 28.5 GeV in the SLAC linac with a bunch intensity of 1-2*10/sup 10/. The beam was delivered to the Final Focus Test Beam Facility (FFTB) at a rate of 1 or 10 Hz. At the focal point of the FFTB transport, a special plasma chamber contains a 3 mm diameter pulsed gas nozzle through which either hydrogen or nitrogen gas is "puffed" into the ultrahigh vacuum system at plenum gas pressures up to 75 atm with a discharge time of 800 mu s. The gas is pumped off by a Roots-type pump. On either side of the central chamber are differential pumping sections semi- isolated from each other by thin titanium windows with small (2-5 mm diameter) apertures for the positron beams to pass through. These sections are evacu...

  5. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    International Nuclear Information System (INIS)

    McDeavitt, Sean; Shao, Lin; Tsvetkov, Pavel; Wirth, Brian; Kennedy, Rory

    2014-01-01

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  6. Fuel Performance Experiments and Modeling: Fission Gas Bubble Nucleation and Growth in Alloy Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean [Texas A & M Univ., College Station, TX (United States); Shao, Lin [Texas A & M Univ., College Station, TX (United States); Tsvetkov, Pavel [Texas A & M Univ., College Station, TX (United States); Wirth, Brian [Univ. of Tennessee, Knoxville, TN (United States); Kennedy, Rory [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-04-07

    Advanced fast reactor systems being developed under the DOE's Advanced Fuel Cycle Initiative are designed to destroy TRU isotopes generated in existing and future nuclear energy systems. Over the past 40 years, multiple experiments and demonstrations have been completed using U-Zr, U-Pu-Zr, U-Mo and other metal alloys. As a result, multiple empirical and semi-empirical relationships have been established to develop empirical performance modeling codes. Many mechanistic questions about fission as mobility, bubble coalescience, and gas release have been answered through industrial experience, research, and empirical understanding. The advent of modern computational materials science, however, opens new doors of development such that physics-based multi-scale models may be developed to enable a new generation of predictive fuel performance codes that are not limited by empiricism.

  7. Unconventional gas experience at El Paso Production Company : tapping into deep, tight gas and coalbed methane

    International Nuclear Information System (INIS)

    Bartley, R.L.

    2003-01-01

    The current conditions in the natural gas industry were reviewed, from the excellent current and projected energy prices to low activity and rig count. Various graphs were presented, depicting total proved dry gas reserves and annual production over time for the Gulf of Mexico, including its continental shelf, the Texas coastal plains, and the United States lower 48. Offshore growth of unconventional gas was also displayed. The key elements of the strategy were also discussed. These included: (1) earnings driven, (2) superior science, (3) innovative application of technology, (4) ability to act quickly and decisively, (5) leadership, management, and professional development, and (6) achieve learning curve economics. The core competencies were outlined along with recent discoveries in South Texas and the Upper Gulf Coast. figs

  8. Flue gas injection into gas hydrate reservoirs for methane recovery and carbon dioxide sequestration

    International Nuclear Information System (INIS)

    Yang, Jinhai; Okwananke, Anthony; Tohidi, Bahman; Chuvilin, Evgeny; Maerle, Kirill; Istomin, Vladimir; Bukhanov, Boris; Cheremisin, Alexey

    2017-01-01

    Highlights: • Flue gas was injected for both methane recovery and carbon dioxide sequestration. • Kinetics of methane recovery and carbon dioxide sequestration was investigated. • Methane-rich gas mixtures can be produced inside methane hydrate stability zones. • Up to 70 mol% of carbon dioxide in the flue gas was sequestered as hydrates. - Abstract: Flue gas injection into methane hydrate-bearing sediments was experimentally investigated to explore the potential both for methane recovery from gas hydrate reservoirs and for direct capture and sequestration of carbon dioxide from flue gas as carbon dioxide hydrate. A simulated flue gas from coal-fired power plants composed of 14.6 mol% carbon dioxide and 85.4 mol% nitrogen was injected into a silica sand pack containing different saturations of methane hydrate. The experiments were conducted at typical gas hydrate reservoir conditions from 273.3 to 284.2 K and from 4.2 to 13.8 MPa. Results of the experiments show that injection of the flue gas leads to significant dissociation of the methane hydrate by shifting the methane hydrate stability zone, resulting in around 50 mol% methane in the vapour phase at the experimental conditions. Further depressurisation of the system to pressures well above the methane hydrate dissociation pressure generated methane-rich gas mixtures with up to 80 mol% methane. Meanwhile, carbon dioxide hydrate and carbon dioxide-mixed hydrates were formed while the methane hydrate was dissociating. Up to 70% of the carbon dioxide in the flue gas was converted into hydrates and retained in the silica sand pack.

  9. Shale-Gas Experience as an Analog for Potential Wellbore Integrity Issues in CO2 Sequestration

    Energy Technology Data Exchange (ETDEWEB)

    Carey, James W. [Los Alamos National Laboratory; Simpson, Wendy S. [Los Alamos National Laboratory; Ziock, Hans-Joachim [Los Alamos National Laboratory

    2011-01-01

    Shale-gas development in Pennsylvania since 2003 has resulted in about 19 documented cases of methane migration from the deep subsurface (7,0000) to drinking water aquifers, soils, domestic water wells, and buildings, including one explosion. In all documented cases, the methane leakage was due to inadequate wellbore integrity, possibly aggravated by hydrofracking. The leakage of methane is instructive on the potential for CO{sub 2} leakage from sequestration operations. Although there are important differences between the two systems, both involve migrating, buoyant gas with wells being a primary leakage pathway. The shale-gas experience demonstrates that gas migration from faulty wells can be rapid and can have significant impacts on water quality and human health and safety. Approximately 1.4% of the 2,200 wells drilled into Pennsylvania's Marcellus Formation for shale gas have been implicated in methane leakage. These have resulted in damage to over 30 domestic water supplies and have required significant remediation via well repair and homeowner compensation. The majority of the wellbore integrity problems are a result of over-pressurization of the wells, meaning that high-pressure gas has migrated into an improperly protected wellbore annulus. The pressurized gas leaks from the wellbore into the shallow subsurface, contaminating drinking water or entering structures. The effects are localized to a few thousands of feet to perhaps two-three miles. The degree of mixing between the drinking water and methane is sufficient that significant chemical impacts are created in terms of elevated Fe and Mn and the formation of black precipitates (metal sulfides) as well as effervescing in tap water. Thus it appears likely that leaking CO{sub 2} could also result in deteriorated water quality by a similar mixing process. The problems in Pennsylvania highlight the critical importance of obtaining background data on water quality as well as on problems associated with

  10. Peculiarities occurrence and microstrip gas chambers studied through experiment WA97; La production d`etrangete et les chambres gazeuses a micropistes dans le cadre de l`experience WA97

    Energy Technology Data Exchange (ETDEWEB)

    Kachelhoffer, T.

    1995-04-01

    This paper presents the studies on development of a Monte-Carlo type generator used for inclusive production of odd baryons and antibaryons through proton- proton and proton- nucleus collisions. Experiment WA97 consisted in designing simulation software for MSGCs (Micro-strips Gas Chambers) as well as the redefining of particle paths with the help of these chambers. This work made it possible to design the MSGC detector for experiment WA97. (TEC). 71 refs., 88 figs.

  11. Deregulation and natural gas trade relationships: lessons from the Alberta-California experience

    International Nuclear Information System (INIS)

    Wilson, Patrick Impero

    1997-01-01

    In 1978 the US government moved to deregulate the American natural gas industry. The market changes that resulted from this initial step took time to ripple their way out to regional and subnational gas trading relationships. This ripple effect required subnational governments (state and provincial regulators) to rethink their gas regulatory policies. This article examines the restructuring of the Alberta-California gas trade. It explores how changes in US policy forced California and Alberta regulators to recast their policies. It concludes with several lessons that can be drawn from this case about the complex challenge of restructuring international gas trading relationships. (author)

  12. Data report of ROSA/LSTF experiment SB-HL-12. 1% hot leg break LOCA with SG depressurization and gas inflow

    International Nuclear Information System (INIS)

    Takeda, Takeshi

    2016-01-01

    An experiment SB-HL-12 was conducted on February 24, 1998 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-HL-12 simulated a 1% hot leg small-break loss-of-coolant accident in a pressurized water reactor under assumptions of total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). Steam generator (SG) secondary-side depressurization by fully opening the relief valves in both SGs as an accident management (AM) action was initiated immediately after maximum surface temperature of simulated fuel rod reached 600 K. Auxiliary feedwater injection into the secondary-side of both SGs was started immediately after the initiation of AM action. After the onset of AM action due to first core uncovery by core boil-off, the primary pressure decreased following the SG secondary-side pressure, causing core mixture level swell. The fuel rod surface temperature then increased up to 635 K. Second core uncovery by core boil-off took place before loop seal clearing (LSC) induced by steam condensation on ACC coolant injected into cold legs. The core liquid level recovered rapidly after the LSC. The fuel rod surface temperature then increased up to 696 K. The pressure difference became larger between the primary and SG secondary sides after the ACC tanks started to discharge nitrogen gas, which resulted in no actuation of LPI system of ECCS during the experiment. Third core uncovery by core boil-off occurred during the reflux condensation in the SG U-tubes under nitrogen gas inflow. The core power was automatically decreased by the LSTF core protection system when the maximum fuel rod surface temperature exceeded 908 K. The obtained data would be useful to define the conditions for counterpart testing of other integral test facilities to address scaling problems through thermal

  13. Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the Monju Reactor Vessel. Final Report of a Coordinated Research Project 2008-2012

    International Nuclear Information System (INIS)

    2014-11-01

    The IAEA supports Member States in the area of advanced fast reactor technology development by providing a major fulcrum for information exchange and collaborative research programmes. The IAEA’s activities in this field are mainly carried out within the framework of the Technical Working Group on Fast Reactors (TWG-FR), which assists in the implementation of corresponding IAEA support, and ensures that all technical activities are in line with expressed needs of Member States. Among this broad range, the IAEA proposes and establishes coordinated research projects (CRPs), aimed at improving Member State capability in fast reactor design and analysis. An important opportunity to perform collaborative research activities was provided by the system startup tests carried out by the Japan Atomic Energy Agency (JAEA) in the prototype loop type sodium cooled fast reactor Monju, in particular a turbine trip test performed in December 1995. As the JAEA opened the experimental dataset to international collaboration in 2008, the IAEA launched the CRP on Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the Monju Reactor Vessel. The CRP, together with eight institutes from seven States, has contributed to improving capabilities in sodium cooled fast reactors simulation through code verification and validation, with particular emphasis on thermal stratification and natural circulation phenomena

  14. Gas supply and Yorkshire Electricity

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1995-04-01

    Yorkshire Electricity, among other independent suppliers of gas, now competes for a share of the United Kingdom gas market, previously monopolised by British Gas. The experience of this successful electric utility company, expanding into the industrial and domestic gas supply market is described in the article. The company`s involvement stems partly from the fact that significant volumes of gas are landed at three terminals within its franchise area. The company will also seek to use subsidaries to generate electric power from gas turbine power plants and explore the possibilities of developing combined heat and power (CHP) plants where appropriate. (UK)

  15. Measurement of air and VOC vapor fluxes during gas-driven soil remediation: bench-scale experiments.

    Science.gov (United States)

    Kim, Heonki; Kim, Taeyun; Shin, Seungyeop; Annable, Michael D

    2012-09-04

    In this laboratory study, an experimental method was developed for the quantitative analyses of gas fluxes in soil during advective air flow. One-dimensional column and two- and three-dimensional flow chamber models were used in this study. For the air flux measurement, n-octane vapor was used as a tracer, and it was introduced in the air flow entering the physical models. The tracer (n-octane) in the gas effluent from the models was captured for a finite period of time using a pack of activated carbon, which then was analyzed for the mass of n-octane. The air flux was calculated based on the mass of n-octane captured by the activated carbon and the inflow concentration. The measured air fluxes are in good agreement with the actual values for one- and two-dimensional model experiments. Using both the two- and three-dimensional models, the distribution of the air flux at the soil surface was measured. The distribution of the air flux was found to be affected by the depth of the saturated zone. The flux and flux distribution of a volatile contaminant (perchloroethene) was also measured by using the two-dimensional model. Quantitative information of both air and contaminant flux may be very beneficial for analyzing the performance of gas-driven subsurface remediation processes including soil vapor extraction and air sparging.

  16. Interstellar silicate analogs for grain-surface reaction experiments: Gas-phase condensation and characterization of the silicate dust grains

    Energy Technology Data Exchange (ETDEWEB)

    Sabri, T.; Jäger, C. [Laboratory Astrophysics Group of the Max Planck Institute for Astronomy at the Friedrich Schiller University Jena Institute of Solid State Physics, Helmholtzweg 3, D-07743 Jena (Germany); Gavilan, L.; Lemaire, J. L.; Vidali, G. [Observatoire de Paris/Université de Cergy-Pontoise, 5 mail Gay Lussac, F-95000 Cergy-Pontoise (France); Mutschke, H. [Laboratory Astrophysics Group of the Astrophysical Institute and University Observatory, Friedrich Schiller University Jena Schillergässchen 3, D-07743 Jena (Germany); Henning, T., E-mail: tolou.sabri@uni-jena.de [Max Planck Institute for Astronomy Königstuhl 17, D-69117 Heidelberg (Germany)

    2014-01-10

    Amorphous, astrophysically relevant silicates were prepared by laser ablation of siliceous targets and subsequent quenching of the evaporated atoms and clusters in a helium/oxygen gas atmosphere. The described gas-phase condensation method can be used to synthesize homogeneous and astrophysically relevant silicates with different compositions ranging from nonstoichiometric magnesium iron silicates to pyroxene- and olivine-type stoichiometry. Analytical tools have been used to characterize the morphology, composition, and spectral properties of the condensates. The nanometer-sized silicate condensates represent a new family of cosmic dust analogs that can generally be used for laboratory studies of cosmic processes related to condensation, processing, and destruction of cosmic dust in different astrophysical environments. The well-characterized silicates comprising amorphous Mg{sub 2}SiO{sub 4} and Fe{sub 2}SiO{sub 4}, as well as the corresponding crystalline silicates forsterite and fayalite, produced by thermal annealing of the amorphous condensates, have been used as real grain surfaces for H{sub 2} formation experiments. A specifically developed ultra-high vacuum apparatus has been used for the investigation of molecule formation experiments. The results of these molecular formation experiments on differently structured Mg{sub 2}SiO{sub 4} and Fe{sub 2}SiO{sub 4} described in this paper will be the topic of the next paper of this series.

  17. Irradiation experiments of 3rd, 4th and 5th fuel assemblies by an in-pile gas loop, OGL-1

    International Nuclear Information System (INIS)

    Fukuda, Kousaku; Kobayashi, Fumiaki; Hayashi, Kimio; Minato, Kazuo; Kikuchi, Teruo; Adachi, Mamoru; Iwamoto, Kazumi; Ikawa, Katsuichi; Itami, Hiroharu.

    1986-07-01

    Three irradiation experiments for 3rd, 4th and 5th fuel assemblies which had been composed of VHTR reference coated particle fuels and graphite components were carried out by an in-pile gas loop, OGL-1 during 1979 and 1982. The main purposes of these experiments were to study on bowing of the fuel rod by irradiation for the 3rd fuel assembly, to study on fuel behavior under relatively low burnup irradiation for the 4th fuel assembly, and to study on fuel behavior up to full burnup of VHTR design for the 5th fuel assembly. For understanding in-pile fuel behavior, fractional releases of fission gases from each fuel assembly were estimated by measuring the fission gas concentrations in the primary loop of OGL-1. The post-irradiation examination (PIE) was carried out extensively on the fuel block, the fuel rods and the fuel compacts in Tokai Hot Laboratory. Also, made were the measurements of metallic fission product distributions in the fuel assemblies and the fuel rods. The results in these experiments were given as follows ; bowing of the fuel rod in the 3rd fuel assembly was 0.7 mm, but integrity of the rod was kept under irradiation. Fractional release of the fission gas from the 4th fuel assembly remained in the order of 10 -7 during irradiation, suggesting that the fuel performance was excellent. The fractional release from the 5th fuel assembly, on the other hand, was in the order of 10 -5 which was the same level in the VHTR design. (author)

  18. IR and UV gas absorption measurements during NOx reduction on an industrial natural gas fired power plant

    DEFF Research Database (Denmark)

    Stamate, Eugen; Chen, Weifeng; Jørgensen, L.

    2010-01-01

    NOx reduction of flue gas by plasma-generated ozone was investigated in pilot test experiments on an industrial power plant running on natural gas. Reduction rates higher than 95% have been achieved for a molar ratio O3:NOx slightly below two. Fourier transform infrared and ultraviolet absorption...... spectroscopy were used for spatial measurements of stable molecules and radicals along the reduction reactor. Reactions of O3 injected in the flue gas in the reduction reactor were also modeled. Experiments are in good agreement with numerical simulations. The operation costs for NOx reduction were estimated...

  19. PBDOWN: A computer code for simulation of core material discharge and expansion in the upper coolant plenum in a hypothetical unprotected loss of flow accident in a LMFBR

    International Nuclear Information System (INIS)

    Royl, P.

    1985-01-01

    The report gives a description of the code PBDOWN (Pool Blow Down), its equations, input specifications and subroutines and it lists the input and output for some samples. Besides that some analysis results for the SNR-300 are discussed, that were obtained with this code. PBDOWN is an integral blow-down and expansion code, which simulates core material discharge and expansion into a sodium filled upper coolant plenum after build-up of vapour pressures in an unprotected loss of flow accident. The model includes the effect of sodium entrainment into an expending bubble of fuel or steel vapour with various assumptions for the heat transfer and vaporization of the entrained sodium droplets. The expanding vapour bubble is connected to the discharging pool via an orifice of a given size through which a time dependent ejection is simulated using quasi-stationary blow down correlations. The model allows bounding analysis of the possible influence of sodium vapour as a secondary working fluid, that is activated outside the pool on the overall expansion energy and discharge

  20. Cogeneration with natural gas fired internal combustion engines: Italian utility's 10 years operating experience

    International Nuclear Information System (INIS)

    Montermini, G.P.

    1992-01-01

    This paper describes the experience that AGAC, an Italian gas and water utility, has acquired in the operation of a 116 Km long district heating network serving about 40,000 inhabitants. The network is powered by a mix of methane fuelled Otto and diesel cycle engines, coal fired fluidized bed boilers, and methane fired boilers producing annually about 153,000 kW of thermal energy, 2,300 kW of cooling energy, and 28.8 million kWh of electric power. This paper reports on the performance of this system in terms of production and sales trends, equipment efficiency and compatibility with new European Communities air pollution standards

  1. Structure of gas pressure signal at two-orifice bubbling from a common plenum

    Czech Academy of Sciences Publication Activity Database

    Růžička, Marek; Drahoš, Jiří; Zahradník, Jindřich; Thomas, N. H.

    2000-01-01

    Roč. 55, č. 2 (2000), s. 421-429 ISSN 0009-2509 R&D Projects: GA ČR GA104/98/1435; GA AV ČR KSK2040602 Grant - others:INCO-COPERNICUS(XE) ERB IC15-CT98-0904 Subject RIV: CI - Industrial Chemistry, Chemical Engineering Impact factor: 1.053, year: 2000

  2. The Effect of Rain on Air-Water Gas Exchange

    Science.gov (United States)

    Ho, David T.; Bliven, Larry F.; Wanninkhof, Rik; Schlosser, Peter

    1997-01-01

    The relationship between gas transfer velocity and rain rate was investigated at NASA's Rain-Sea Interaction Facility (RSIF) using several SF, evasion experiments. During each experiment, a water tank below the rain simulator was supersaturated with SF6, a synthetic gas, and the gas transfer velocities were calculated from the measured decrease in SF6 concentration with time. The results from experiments with IS different rain rates (7 to 10 mm/h) and 1 of 2 drop sizes (2.8 or 4.2 mm diameter) confirm a significant and systematic enhancement of air-water gas exchange by rainfall. The gas transfer velocities derived from our experiment were related to the kinetic energy flux calculated from the rain rate and drop size. The relationship obtained for mono-dropsize rain at the RSIF was extrapolated to natural rain using the kinetic energy flux of natural rain calculated from the Marshall-Palmer raindrop size distribution. Results of laboratory experiments at RSIF were compared to field observations made during a tropical rainstorm in Miami, Florida and show good agreement between laboratory and field data.

  3. Norm waste in oil and gas industry: The Syrian experience

    International Nuclear Information System (INIS)

    Al-Masri, M.S.; Suman, H.

    2001-01-01

    This paper describes the Syrian experience in respect to Naturally Occurring Radioactive Materials (NORM) waste in Syrian oil and gas industry. NORM can be concentrated and accumulated in tubing and surface equipment of oil and gas production lines in the form of scale and sludge. NORM waste (scale, sludge, production water) is therefore generated during cleaning, physical or chemical treatment of streams. Uncontrolled disposal of this type of waste could lead to environmental pollution, and thus eventually to exposure of members of the public. The presence of NORM in Syrian oil fields has been recognized since 1987 and AECS has initiated several studies, in cooperation with oil companies, to manage such type of waste. Three categories of NORM waste in Syrian oil fields were identified. Firstly, hard scales from either decontamination of contaminated equipment and tubular using high-pressure water systems or mechanical cleaning at site are considered to contain the highest levels of radium isotopes ( 226 Ra, 228 Ra, 224 Ra). Secondly, sludge wastes are generated with large amount but low levels of radium isotopes were found. Thirdly, contaminated soil with 226 Ra as a result of uncontrolled disposal of production water was also considered as NORM waste. The first waste type (scale) is stored in Standard storage barrels in a controlled area; the number of barrels is increasing with time. High levels of radium isotopes were found in these scales. The options for disposal of these wastes are still under investigations; one of the most predominant thinking is the re-injection into abundant wells. For sludge waste, plastic lined disposal pits were constructed in each area for temporary storage. Moreover, big gas power stations have been built and operated since the last ten years. Maintenance operations for these stations produce tens of tones of scales containing radon daughters, 210 Pb and 210 Po with relatively high concentrations. The common practice used to dispose

  4. An Advanced Analytical Chemistry Experiment Using Gas Chromatography-Mass Spectrometry, MATLAB, and Chemometrics to Predict Biodiesel Blend Percent Composition

    Science.gov (United States)

    Pierce, Karisa M.; Schale, Stephen P.; Le, Trang M.; Larson, Joel C.

    2011-01-01

    We present a laboratory experiment for an advanced analytical chemistry course where we first focus on the chemometric technique partial least-squares (PLS) analysis applied to one-dimensional (1D) total-ion-current gas chromatography-mass spectrometry (GC-TIC) separations of biodiesel blends. Then, we focus on n-way PLS (n-PLS) applied to…

  5. LHC-GCS Process Tuning selection and use of PID and Smith predictor for the regulations of the LHC experiments' gas systems

    CERN Document Server

    Cabaret, S; Rachid, A; Coppier, H

    2005-01-01

    The LHC experiment’s Gas Control System (LHC GCS) has to provide LHC experiments with homogeneous control systems (supervision and process control layers) for their 23 gas systems. The LHC GCS process control layer is based on Programmable Logic Controllers (PLCs), Field-Buses and on a library, UNICOS (UNified Industrial COntrol System). Its supervision layer is based on a commercial SCADA system and on the JCOP and UNICOS PVSS frameworks. A typical LHC experiment’s gas system is composed of up to ten modules, dedicated to specific functions (e.g. mixing, purification, circulation). Most of modules require control loops for the regulation of pressures, temperatures and flows or ratios of gases. The control loops of the 23 gas systems can be implemented using the same tools, but need specific tuning according to their respective size, volume, pipe lengths and required accuracy. Most of the control loops can be implemented by means a standard PID (Proportional, Integral and Derivative) controller. When this...

  6. Effect of ribbed and smooth coolant cross-flow channel on film cooling

    International Nuclear Information System (INIS)

    Peng, Wei; Sun, Xiaokai; Jiang, Peixue; Wang, Jie

    2017-01-01

    Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

  7. Effect of ribbed and smooth coolant cross-flow channel on film cooling

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei; Sun, Xiaokai [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Jiang, Peixue, E-mail: jiangpx@tsinghua.edu.cn [Key Laboratory for Thermal Science and Power Engineering of Ministry of Educations, Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China); Wang, Jie [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

  8. Slug flooding in air-water countercurrent vertical flow

    International Nuclear Information System (INIS)

    Lee, Jae Young; Raman, Roger; Chang, Jen-Shih

    2000-01-01

    This paper is to study slug flooding in the vertical air-water countercurrent flow loop with a porous liquid injector in the upper plenum. More water penetration into the bottom plenum in slug flooding is observed than the annular flooding because the flow regime changes from the slug flow regime or periodic slug/annular flow regime to annular flow regime due to the hysteresis between the onset of flooding and the bridging film. Experiments were made tubes of 0.995 cm, 2.07 cm, and 5.08 cm in diameter. A mechanistic model for the slug flooding with the solitary wave whose height is four time of the mean film thickness is developed to produce relations of the critical liquid flow rate and the mean film thickness. After fitting the critical liquid flow rate with the experimental data as a function of the Bond number, the gas flow rate for the slug flooding is obtained by substituting the critical liquid flow rate to the annular flooding criteria. The present experimental data evaluate the slug flooding condition developed here by substituting the correlations for mean film thickness models in the literature. The best prediction was made by the correlation for the mean film thickness of the present study which is same as Feind's correlation multiplied by 1.35. (author)

  9. Measurements of bundle end flux peaking effects in 37-element CANDU PHW fuel

    International Nuclear Information System (INIS)

    French, P.M.

    1977-10-01

    Thermal neutron bundle end flux peaking factors have been measured in fresh 37-element Bruce reactor natural UO 2 clusters in heavy water moderator, both with and without staggered plenums at the fuel stack ends, in representative elements throughout the clusters. The measurements were made at a square lattice pitch of 28.58 cm with heavy water coolant. The results indicate that outer element peaking factors are 1.142 +- 0.009 for bundles containing no plenums, and 1.155 +- 0.006 and 1.177 +- 0.006 at the non-plenum and plenum element ends respectively, for bundles containing staggered plenums, irrespective of the azimuthal orientation between pairs of bundles. Measurements are also reported for bundles containing plenums in every outer element, for bundles separated by a stainless steel flux suppressor, for longer graphite plenums, and for changes in plenum and bundle gap lengths. Some theoretical comparisons with the results, reported by other authors, have been summarized. (author)

  10. The design, fabrication and testing of the gas analysis system for the tritium recovery experiment, TRIO-01

    International Nuclear Information System (INIS)

    Finn, P.A.; Bowers, D.L.; Clemmer, E.D.; Clemmer, R.G.; Graczyk, D.G.; Homa, M.I.; Pappas, G.; Reedy, G.T.; Slawecki, M.A.

    1983-01-01

    The tritium recovery experiment, TRIO-01, required a gas analysis system which detected the form of tritium, the amount of tritium (differential and integral), and the presence and amount of other radioactive species. The system had to handle all contingencies and function for months at a time; unattended during weekend operation. The designed system, described herein, consisted of a train of components which could be grouped as desired to match tritium release behavior

  11. Experimental study on engine gas-path component fault monitoring using exhaust gas electrostatic signal

    International Nuclear Information System (INIS)

    Sun, Jianzhong; Zuo, Hongfu; Liu, Pengpeng; Wen, Zhenhua

    2013-01-01

    This paper presents the recent development in engine gas-path components health monitoring using electrostatic sensors in combination with signal-processing techniques. Two ground-based engine electrostatic monitoring experiments are reported and the exhaust gas electrostatic monitoring signal-based fault-detection method is proposed. It is found that the water washing, oil leakage and combustor linear cracking result in an increase in the activity level of the electrostatic monitoring signal, which can be detected by the electrostatic monitoring system. For on-line health monitoring of the gas-path components, a baseline model-based fault-detection method is proposed and the multivariate state estimation technique is used to establish the baseline model for the electrostatic monitoring signal. The method is applied to a data set from a turbo-shaft engine electrostatic monitoring experiment. The results of the case study show that the system with the developed method is capable of detecting the gas-path component fault in an on-line fashion. (paper)

  12. Gas-cooled reactors

    International Nuclear Information System (INIS)

    Schulten, R.; Trauger, D.B.

    1976-01-01

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

  13. Activation calculations for dismantling - The feedback of a 7 years experience in activation calculations for graphite gas cooled reactors in France

    International Nuclear Information System (INIS)

    Eid, M.; Nimal, J.C.; Gerat, L.M.

    1994-01-01

    This is a revision of the past seven years experience in activation calculations for dismantling. It aims at evaluating the experience and at making better understanding to help in decision making during the following phases. Five gas cooled reactors are shutdown and are waiting for the EDF (Electricite De France) dismantling decision. The sixth (BUGEY1) will be shutdown by 1994 and will be waiting a dismantling decision as well. (authors). 3 figs., 3 tabs

  14. GAM - Gas Migration Experiments in a Heterogeneous Shear Zone of the Grimsel Test of the Grimsel Test Site

    International Nuclear Information System (INIS)

    Marschall, P.; Lunati, I.

    2006-12-01

    This report documents the scientific investigations carried out as part of the GAM project between June 1997 and April 2001 at the Grimsel Test Site within the framework of Investigation Phase V (1997 - 2001). Four radioactive waste management organisations participated in the GAM experiment, namely ANDRA, ENRESA, NAGRA and Sandia National Laboratories for the US Department of Energy. The experiment team consisted of the delegates of the participating organisations, research groups from the Swiss Federal Institute of Technology, Zurich and from the Technical University of Catalonia, Barcelona and, last but not least, several contractor teams. Essential aims of the GAM investigation programme were the development and testing of laboratory and field equipment for tracer experiments. Innovative laboratory technologies were applied, such as Laser Scanning Confocal Microscopy and X-ray tomography, flow visualisation in artificial fractures, nuclear magnetic resonance measurements and neutron radiography. Furthermore, a new technique was tested for the recovery of well preserved core samples from the GAM shear zone. Novelties in field testing comprised the use of an on-line counter for the particle tracer tests and a georadar survey of gas and brine injection tests with a high frequency borehole antenna. The development of upscaling methodologies and the derivation of effective parameters for single- and two-phase flow models was another issue of interest. The investigations comprised theoretical studies on solute transport in non-uniform flow fields and assessment of the impact of the microstructure on solute and gas transport. Closely related to these theoretical studies was the numerical interpretation of the combined solute and gas tracer tests, which revealed the great potential of such data sets with regard to model discrimination. As a final step in the synthesis task of the GAM project, a model abstraction process was established, aimed at integrating the

  15. On heat transfer characteristics of real and simulant melt pool experiments

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Nourgaliev R.R.; Sehgal, B.R. [Royal Institute of Technology, Stockholm (Sweden)

    1995-09-01

    The paper presents results of analytical studies of natural convection heat transfer in scaled and/or simulant melt pool experiments related to the PWR in-vessel melt retention issue. Specific reactor-scale effects of a large decay-heated core melt pool in the reactor pressure vessel lower plenum are first reviewed, and then the current analytical capability of describing physical processes under prototypical situations is examined. Experiments and experimental approaches are analysed by focusing on their ability to represent prototypical situations. Calculations are carried out in order to assess the significance of some selected effects, including variations in melt properties, pool geometry and heating conditions. Rayleigh numbers in the present analysis are limited to 10{sup 12}, where uncertainties in turbulence modeling are not overriding other uncertainties. The effects of fluid Prandtl number on heat transfer to the lowermost part of cooled pool walls are examined for square and semicircular cavities. Calculations are performed also to explore limitations of using side-wall heating and direct electrical heating in reproducing the physical picture of interest. Needs for further experimental and analytical efforts are discussed as well.

  16. Exobiological implications of dust aggregation in planetary atmospheres: An experiment for the gas-grain simulation facility

    Science.gov (United States)

    Huntington, J. L.; Schwartz, D. E.; Marshall, J. R.

    1991-01-01

    The Gas-Grain Simulation Facility (GGSF) will provide a microgravity environment where undesirable environmental effects are reduced, and thus, experiments involving interactions between small particles and grains can be more suitably performed. Slated for flight aboard the Shuttle in 1992, the ESA glovebox will serve as a scientific and technological testbed for GGSF exobiology experiments as well as generating some basic scientific data. Initial glovebox experiments will test a method of generating a stable, mono-dispersed cloud of fine particles using a vibrating sprinkler system. In the absence of gravity and atmospheric turbulence, it will be possible to determine the influence of interparticle forces in controlling the rate and mode of aggregation. The experimental chamber can be purged of suspended matter to enable multiple repetitions of the experiments. Of particular interest will be the number of particles per unit volume of the chamber, because it is suspected that aggregation will occur extremely rapidly if the number exceeds a critical value. All aggregation events will be recorded on high-resolution video film. Changes in the experimental procedure as a result of surprise events will be accompanied by real-time interaction with the mission specialist during the Shuttle flight.

  17. A method for measuring the local gas pressure within a gas-flow stage in situ in the transmission electron microscope

    International Nuclear Information System (INIS)

    Colby, R.; Alsem, D.H.; Liyu, A.; Kabius, B.

    2015-01-01

    Environmental transmission electron microscopy (TEM) has enabled in situ experiments in a gaseous environment with high resolution imaging and spectroscopy. Addressing scientific challenges in areas such as catalysis, corrosion, and geochemistry can require pressures much higher than the ∼20 mbar achievable with a differentially pumped environmental TEM. Gas flow stages, in which the environment is contained between two semi-transparent thin membrane windows, have been demonstrated at pressures of several atmospheres. However, the relationship between the pressure at the sample and the pressure drop across the system is not clear for some geometries. We demonstrate a method for measuring the gas pressure at the sample by measuring the ratio of elastic to inelastic scattering and the defocus of the pair of thin windows. This method requires two energy filtered high-resolution TEM images that can be performed during an ongoing experiment, at the region of interest. The approach is demonstrated to measure greater than atmosphere pressures of N 2 gas using a commercially available gas-flow stage. This technique provides a means to ensure reproducible sample pressures between different experiments, and even between very differently designed gas-flow stages. - Highlights: • Method developed for measuring gas pressure within a gas-flow stage in the TEM. • EFTEM and CTF-fitting used to calculate amount and volume of gas. • Requires only a pair of images without leaving region of interest. • Demonstrated for P > 1 atm with a common commercial gas-flow stage

  18. Gas-phase fragmentation of peptides to increase the spatial resolution of the Hydrogen Exchange Mass Spectrometry experiment

    DEFF Research Database (Denmark)

    Jensen, Pernille Foged; Rand, Kasper Dyrberg

    2016-01-01

    are produced after precursor ion selection and thus do not add complexity to the LC-MS analysis. The key to obtaining optimal spatial resolution in a hydrogen exchange mass spectrometry (HX-MS) experiment is the fragmentation efficiency. This chapter discusses common fragmentation techniques like collision....../D scrambling, thus making them suitable for HX applications. By combining the classic bottom-up HX-MS workflow with gas-phase fragmentation by ETD, detailed information on protein HX can be obtained....

  19. The gas introduction system of JET

    International Nuclear Information System (INIS)

    Boschi, A.; Dietz, K.J.; Rebut, P.H.

    1984-01-01

    The Gas Introduction System of JET is designed to handle, measure, transfer and inject into the machine, at given rates and times, the quantities of gases required to feel the plasma discharges. The System is composed by a Gas Handling Unit for the gas preparation, and four identical Gas Introduction Modules which are positioned symmetrically at the machine. The lay-out and design of the different components is described and operational experience is presented. (author)

  20. The Nigerian experience in health, safety, and environmental matters during oil and gas exploration and production operations

    International Nuclear Information System (INIS)

    Oyekan, A.J.

    1991-01-01

    Since crude oil was first discovered in commercial quantities in the Country, in 1956, Nigerian oil and gas exploration and production activities have steadily increased as petroleum assumed strategic importance in the nation's economy. However, just as occurs in many parts of the world, crude oil and gas are found and produced in Nigeria sometimes in very hostile and unfavorable environments. The search for oil and gas takes explorers to the hot regions of the Northern parts of the country, the swamp jungle location of the Niger Delta, as well as offshore locations in the Atlantic Ocean. Each terrain, whether land, swamp or offshore, in deep or shallow waters, present unique health, safety and environmental implications and challenges to the operators, as well as, to the Government regulators. From a background of existing Nigerian Laws and operational experience, this paper details the programmes that have been put in place to guarantee a healthy workforce, ensure the safety of personnel and equipment, and protect the Nigerian environment during oil and gas exploration and production operations, as well as their documented effectiveness. The paper discusses the performance of the Petroleum Industry by analyzing the health, safety and environmental records available from 1956 - 1990. The records of major incidents related to safety and environment over the period are discussed and evaluated. The paper notes that relatively speaking, in spite of the Bomu 2 and Funiwa V oil well blow-outs in 1970 and 1980 respectively which caused extensive environmental damages and the Anieze, Oniku and KC 1 gas well blow-out of 1972, 1975 and 1989 respectively, which resulted in the loss of the rigs drilling the locations concerned, the safety performance records in the Nigerian oil and gas exploration and production activities in the past thirty-five years have been satisfactory compared with the records of similar operations in most other parts of the world

  1. Mitigating greenhouse gas emissions of the agriculture sector in France. Collection of territorial experiences

    International Nuclear Information System (INIS)

    Pommier, Fabien; Martin, Sarah; Bajeat, Philippe; Larbodiere, Ludovic; Vergez, Antonin

    2013-06-01

    After having briefly indicated the different origins of direct and indirect greenhouse gas emissions by the agriculture sector, presented the technical and political context, and outlined the need for new practices to struggle against climate change and to adapt to changes to come, this publication reports some experiments undertaken in different French regions: a farm network as an animation tool to support farmers, a local partnership to conciliate town and agriculture, the development of actions on energy and greenhouse gases in agriculture, the implementation of climate and agriculture plan, a network of agricultural actors for a sustainable support of change, an agriculture with and for its territory and inhabitants, a debate on agriculture and climate

  2. Effects of injection nozzle exit width on rotating detonation engine

    Science.gov (United States)

    Sun, Jian; Zhou, Jin; Liu, Shijie; Lin, Zhiyong; Cai, Jianhua

    2017-11-01

    A series of numerical simulations of RDE modeling real injection nozzles with different exit widths are performed in this paper. The effects of nozzle exit width on chamber inlet state, plenum flowfield and detonation propagation are analyzed. The results are compared with that using an ideal injection model. Although the ideal injection model is a good approximation method to model RDE inlet, the two-dimensional effects of real nozzles are ignored in the ideal injection model so that some complicated phenomena such as the reflected waves caused by the nozzle walls and the reversed flow into the nozzles can not be modeled accurately. Additionally, the ideal injection model overpredicts the block ratio. In all the cases that stabilize at one-wave mode, the block ratio increases as the nozzle exit width gets smaller. The dual-wave mode case also has a relatively high block ratio. A pressure oscillation in the plenum with the same main frequency with the rotating detonation wave is observed. A parameter σ is applied to describe the non-uniformity in the plenum. σ increases as the nozzle exit width gets larger. Under some condition, the heat release on the interface of fresh premixed gas layer and detonation products can be strong enough to induce a new detonation wave. A spontaneous mode-transition process is observed for the smallest exit width case. Due to the detonation products existing in the premixed gas layer before the detonation wave, the detonation wave will propagate through reactants and products alternately, and therefore its strength will vary with time, especially near the chamber inlet. This tendency gets weaker as the injection nozzle exit width increases.

  3. A review of helium gas turbine technology for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    No, Hee Cheon; Kim, Ji Hwan; Kim, Hyeun Min

    2007-01-01

    Current High-Temperature Gas-cooled Reactors (HTGRs) are based on a closed brayton cycle with helium gas as the working fluid. Thermodynamic performance of the axial-flow helium gas turbines is of critical concern as it considerably affects the overall cycle efficiency. Helium gas turbines pose some design challenges compared to steam or air turbomachinery because of the physical properties of helium and the uniqueness of the operating conditions at high pressure with low pressure ratio. This report present a review of the helium Brayton cycle experiences in Germany and in Japan. The design and availability of helium gas turbines for HTGR are also presented in this study. We have developed a new throughflow calculation code to calculate the design-point performance of helium gas turbines. Use of the method has been illustrated by applying it to the GTHTR300 reference

  4. Fixed target flammable gas upgrades

    International Nuclear Information System (INIS)

    Schmitt, R.; Squires, B.; Gasteyer, T.; Richardson, R.

    1996-12-01

    In the past, fixed target flammable gas systems were not supported in an organized fashion. The Research Division, Mechanical Support Department began to support these gas systems for the 1995 run. This technical memo describes the new approach being used to supply chamber gasses to fixed target experiments at Fermilab. It describes the engineering design features, system safety, system documentation and performance results. Gas mixtures provide the medium for electron detection in proportional and drift chambers. Usually a mixture of a noble gas and a polyatomic quenching gas is used. Sometimes a small amount of electronegative gas is added as well. The mixture required is a function of the specific chamber design, including working voltage, gain requirements, high rate capability, aging and others. For the 1995 fixed target run all the experiments requested once through gas systems. We obtained a summary of problems from the 1990 fixed target run and made a summary of the operations logbook entries from the 1991 run. These summaries primarily include problems involving flammable gas alarms, but also include incidents where Operations was involved or informed. Usually contamination issues were dealt with by the experimenters. The summaries are attached. We discussed past operational issues with the experimenters involved. There were numerous incidents of drift chamber failure where contaminated gas was suspect. However analyses of the gas at the time usually did not show any particular problems. This could have been because the analysis did not look for the troublesome component, the contaminant was concentrated in the gas over the liquid and vented before the sample was taken, or that contaminants were drawn into the chambers directly through leaks or sub-atmospheric pressures. After some study we were unable to determine specific causes of past contamination problems, although in argon-ethane systems the problems were due to the ethane only

  5. Dissociation behavior of methane gas hydrate in porous media

    Energy Technology Data Exchange (ETDEWEB)

    Qiang, C.; Yu-gang, Y.; Chang-ling, L. [Ministry of Land and Resources, Quindao (China). Qingdao Inst. of Marine Geology; Qing-guo, M. [Qingdao Univ. College of Chemical Engineering and Environment, Shandong, Qingdao (China)

    2008-07-01

    Gas hydrates are ice-like compounds that form by natural gas and water and are considered to be a new energy resource. In order to make good use of this resource, it is important to know the hydrate dissociation process. This paper discussed an investigation of methane hydrate dissociation through a simulation experiment. The paper discussed the gas hydrates dissociation experiment including the apparatus and experiment equipment, including methane gas supply; reaction cell; temperature controller; pressure maintainer; and gas flow meter. The paper also presented the method and material including iso-volumetric dissociation and normal pressure dissociation. Last, results and discussion of the results were presented. A comparison of five different particle sizes did not reveal any obvious effects that were related to the porous media, mostly likely because the particle size was too large. 15 refs., 2 tabs., 4 figs.

  6. Online gas analysis and diagnosis for RPC detectors in the ATLAS experiment

    International Nuclear Information System (INIS)

    De Asmundis, Riccardo

    2007-01-01

    Resistive Plate Counters (RPC) detectors need a very strict control of gas parameters: motivations for this statement come from both the request of stability in the detector working point, and chemical consideration concerning potentially aggressive materials generated during the ionization processes into the sensitive gap; the latter point can be relevant because of a possible damage to the internal surface of the detector that has to be avoided in order to ensure an high detection efficiency of the RPC during their ten years or more of operation in ATLAS. In order to understand these aspects, detailed studies on gas behavior have been carried on at the GIF-X5 at CERN (2002-2005), based on Gas Chromatographic and spectroscopy techniques. Main results of these analysis are presented here, together with the design of the online analyzer to be installed on ATLAS conceived to keep control of gas quality and to trigger maintenance interventions on the gas system, in particular on the purification subsystem

  7. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 4. Numerical analysis of 1/10 scaled water experiment with the AQUA code

    International Nuclear Information System (INIS)

    Muramatu, Toshiharu; Yamaguchi, Akira

    2004-01-01

    A large-scale sodium-cooled fast breeder reactor in the feasibility studies on commercialized fast reactors has a feature of consideration of thorough simplified and compacted systems and components design to realize drastic economical improvements. Therefore, special attentions should be paid to thermohydraulic designs for gas entrainment behavior from free surface, flow-induced vibration of in-vessel components, thermal stratification in the plenum, thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. A numerical analysis was carried out with a multi-dimensional code AQUA to confirm an applicability to the evaluations for the in-vessel thermohydraulic phenomena using a 1/10 scaled water experiment simulating the large-scale fast breeder reactor in the feasibility studies. From the analysis, the following results were obtained. (1) In-vessel thermohydraulics characterized by a radiated flow pattern to the reactor vessel wall and a strong upward flow through a slit of the upper core structures were evaluated. These characteristics agreed approximately with the water experiment. (2) The upward velocity values at the slit agreed well with the experimental data under a condition of γ z = 0.3 and ξ z = 0.5, though overall evaluations of the in-vessel thermohydraulics were failed to predict quantitatively. (3) The AQUA code is applicable to the in-vessel thermohydraulics evaluations in the feasibility studies, though it is necessary to make further modifications of the calculational models for accurate evaluations. On the one hand, it was confirmed that calculated results for the 1/10 water experimental model and the 1/1 actual-scaled model agreed quantitatively for the in-vessel thermohydraulics characteristics indicated above. (author)

  8. EMC3-EIRENE modeling of toroidally-localized divertor gas injection experiments on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Lore, J.D., E-mail: lorejd@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Reinke, M.L. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); LaBombard, B. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Lipschultz, B. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Churchill, R.M. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Pitts, R.A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Feng, Y. [Max Planck Institute for Plasma Physics, Greifswald (Germany)

    2015-08-15

    Experiments on Alcator C-Mod with toroidally and poloidally localized divertor nitrogen injection have been modeled using the three-dimensional edge transport code EMC3-EIRENE to elucidate the mechanisms driving measured toroidal asymmetries. In these experiments five toroidally distributed gas injectors in the private flux region were sequentially activated in separate discharges resulting in clear evidence of toroidal asymmetries in radiated power and nitrogen line emission as well as a ∼50% toroidal modulation in electron pressure at the divertor target. The pressure modulation is qualitatively reproduced by the modeling, with the simulation yielding a toroidal asymmetry in the heat flow to the outer strike point. Toroidal variation in impurity line emission is qualitatively matched in the scrape-off layer above the strike point, however kinetic corrections and cross-field drifts are likely required to quantitatively reproduce impurity behavior in the private flux region and electron temperatures and densities directly in front of the target.

  9. TRIO experiment

    International Nuclear Information System (INIS)

    Clemmer, R.G.; Finn, P.A.; Malecha, R.F.

    1984-09-01

    The TRIO experiment is a test of in-situ tritium recovery and heat transfer performance of a miniaturized solid breeder blanket assembly. The assembly (capsule) was monitored for temperature and neutron flux profiles during irradiation and a sweep gas flowed through the capsule to an anaytical train wherein the amounts of tritium in its various chemical forms were determined. The capsule was designed to operate at different temperatures and sweep gas conditions. At the end of the experiment the amount of tritium retained in the solid was at a concentration of less than 0.1 wppM. More than 99.9% of tritium generated during the experiment was successfully recovered. The results of the experiment showed that the tritium inventories at the beginning and at the end of the experiment follow a relationship which appears to be characteristic of intragranular diffusion

  10. The Relation between Gas Flow and Combustibility using Actual Engine (Basic Experiment of Gas Flow and Combustibility under Low Load Condition)

    OpenAIRE

    田坂, 英紀; 泉, 立哉; 木村, 正寿

    2003-01-01

    Abstract ###Consideration of the global environment problems by exhaust gas is becoming important in recent years. ###Especially about internal combustion engine, social demand has been increasing about low pollution, high ###efficiency and so on. Controlling gas flow in cylinder becomes the key getting good combustion state in ###various driving states. ###The purpose of the research is analysis about the relation between gas flow and combustibility in the cylinder. ###So we measured gas flo...

  11. The gas introduction system of JET

    International Nuclear Information System (INIS)

    Boschi, A.; Dietz, K.J.; Rebut, P.H.

    1985-01-01

    The Gas Introduction System of JET is designed to handle, measure, transfer and inject into the machine, at given rates and times, the quantitites of gases required to feed the plasma discharges. The System is composed by a Gas Handling Unit for the gas preparation, and four identical Gas Introduction Modules which are positioned symmetrically at the machine. In this paper the lay-out and design of the different components is described and operational experience is presented

  12. Tokamak experiments on JIPP T-II with pulsed gas injection

    International Nuclear Information System (INIS)

    Toi, K.; Itoh, S.; Fujita, J.; Kadota, K.; Kawahata, K.

    1978-02-01

    The confinement of tokamak plasma has been investigated in the wide range of electron density average n sub(e) from 1 x 10 13 to 5 x 10 13 cm -3 by using the pulsed gas injection. The gross energy confinement time increases with increase of electron density and reaches 14 msec. The averaged effective ionic charge derived from plasma conductivity = is about 1 to 2 in the regime of small streaming parameter ( = 0.01 -- 0.08). The ratio of ion temperature to electron one is in the range greater than 0.5. This fact means that the ion energy confinement time is greater than the electron-ion energy relaxation time. Excessive injection of cold neutral gas excites m = 2 MHD oscillations. Much more gas injection leads to the remarkable cooling of plasma periphery and disruptive instabilities. These MHD oscillations and disruptive instabilities have been suppressed by the heating of plasma periphery with the second rapid rise of plasma current. (auth.)

  13. Natural gas hydrates. Experimental techniques and their applications

    Energy Technology Data Exchange (ETDEWEB)

    Ye, Yuguang; Liu, Changling (eds.) [Qingdao Institute of Marine Geology (China). Gas Hydrate Laboratory

    2013-07-01

    Focuses on gas hydrate experiment in laboratory. Intends to provide practical significant parameters for gas hydrate exploration and exploitation in the oceanic and permafrost environments. Consists of different themes that present up-to-date information on hydrate experiments. ''Natural Gas Hydrates: Experimental Techniques and Their Applications'' attempts to broadly integrate the most recent knowledge in the fields of hydrate experimental techniques in the laboratory. The book examines various experimental techniques in order to provide useful parameters for gas hydrate exploration and exploitation. It provides experimental techniques for gas hydrates, including the detection techniques, the thermo-physical properties, permeability and mechanical properties, geochemical abnormalities, stability and dissociation kinetics, exploitation conditions, as well as modern measurement technologies etc.

  14. Experiment data report for Semiscale Mod-1 Test S-05-5 (alternate ECC injection test)

    International Nuclear Information System (INIS)

    Collins, B.L.; Patton, M.L. Jr.; Sackett, K.E.

    1977-04-01

    Recorded test data are presented for Test S-05-5 of the Semiscale Mod-1 alternate ECC injection test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-05-5 was conducted from initial conditions of 2263 psia and 537 0 F to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the cold leg broken loop piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. The upper plenum was vented through a reflood bypass line interconnecting the hot and cold legs of the broken loop

  15. Program plan for correction of US instrument degradation or failure in the Upper Plenum Test Facility (UPTF) in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Rhee, G.S.; Chen, Y.S.; Shotkin, L.M.

    1987-07-01

    This report documents, as of September, 1986, the investigation of the failure or degradation of some of the advanced two-phase flow instruments supplied by the United States Nuclear Regulatory Commission (USNRC) to the German Upper Plenum Test Facility (UPTF). These instruments include Tie-Plate Drag Bodies (DBs), Breakthrough Detectors (BTDs), Loop Drag Disc (DD) paddles, Fluid Distribution Grid (FDG) sensors, and Liquid Level Detector (LLD) sensors. The exact causes for these instrument degradations or failures are not known, but several potential causes have been identified. For DBs and BTDs, the primary mechanism for the degradation appears to be a leakage in the Inconel 600 strain gage encapsulation and the subsequent burnout of the strain gage elements. Excessive loads appear to be the cause of the degradation or failure of the drag discs. The degradation cause for most of the FDGs and LLDs may be either steam/water erosion or mechanical abrasion of the sapphire sensor tips. However, some of the FDG tips were found to be cracked also. The corrective actions are being directed towards identification of the primary causes for the instrument degradation or failure and methods of preventing recurrance and toward minimizing the impact on the test program. All possible action items are being reviewed to arrange them in terms of priority and the likelihood of success so that the best results can be obtained under the constraints of a fixed amount of resources and limited time

  16. Transient behavior of natural circulation for boiling two-phase flow, 2

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Chiang, Jing-Hsien; Mori, Michitugu.

    1991-01-01

    In this set of experiments, natural circulation in boiling two-phase flow has been investigated for power transients, simulating the start-up process in a natural circulation BWR. This was done in order to understand the underlying mechanism of thermo-hydraulic instability which may appear during a start-up. In this paper, geysering is dealt with especially and the driving mechanism is clarified by investigating the stability related to effects of inlet velocity, subcooling, temperature in an outlet plenum and non-heated length between heated section and the outlet plenum. Furthermore, by considering these results and the operational experience in the Dodewaard reactor, recommendations on how the thermo-hydraulic instabilities can be prevented from occurring are proposed concerning a reactor configuration and start-up procedure for natural circulation BWRs. (author)

  17. Use of CFD to predict trapped gas excitation as source of vibration and noise in screw compressors

    Science.gov (United States)

    Willie, James

    2017-08-01

    This paper investigates the source of noise in oil free screw compressors mounted on highway trucks and driven by a power take-off (PTO) transmission system. Trapped gas at the discharge side is suggested as possible source of the excitation of low frequency torsional resonance in these compressors that can lead to noise and vibration. Measurements and lumped mass torsional models have shown low frequency torsional resonance in the drive train of these compressors when they are mounted on trucks. This results in high torque peak at the compressor input shaft and in part to pulsating noise inside the machine. The severity of the torque peak depends on the amplitude of the input torque fluctuation from the drive (electric motor or truck engine). This in turn depends on the prop-shaft angle. However, the source of the excitation of this low torsional resonance inside the machine is unknown. Using CFD with mesh motion at every 1° rotation of the rotors, it is shown that the absence of a pressure equalizing chamber at the discharge can lead to trapped gas creation, which can lead to over-compression, over-heating of the rotors, and to high pressure pulsations at the discharge. Over-compression can lead to shock wave generation at the discharge plenum and the pulsation in pressure can lead to noise generation. In addition, if the frequency of the pressure pulsation in the low frequency range coincides with the first torsional frequency of the drive train the first torsional resonance mode can be excited.

  18. Elements for a European gas policy

    International Nuclear Information System (INIS)

    Darmois, Gilles

    2014-01-01

    This report first discusses the role of gas in the European energy mix, and more particularly how it can be developed in the transport sector to replace oil, in the building and housing sector for heating and hot water production, and in the industry. The author also draws some lessons from the German experience with gas where power plants could not find their place in electricity production whereas lignite mines have been used with a maximum negative impact on the environment. This shows that the share of gas in the energy mix will not increase spontaneously, but will have to take carbon cost into account. Then, the author discusses the potential of shale gas in Europe, comments the first economic and geological lessons learned from the experience of the USA. He comments and discusses the environmental risks related to the exploitation of shale gases, and proposes an overview of European perspectives for gas in terms of reserves, infrastructures, supply and demand. The author outlines the European policy is to be reviewed, notably regarding the value of carbon. Some propositions are finally made regarding gas purchase contract negotiation, gas transport and storage infrastructures in Europe, supply security and strategic storages, investments in energy consumption efficiency in industries as well as for households

  19. Gas generation matrix depletion quality assurance project plan

    International Nuclear Information System (INIS)

    1998-01-01

    The Los Alamos National Laboratory (LANL) is to provide the necessary expertise, experience, equipment and instrumentation, and management structure to: Conduct the matrix depletion experiments using simulated waste for quantifying matrix depletion effects; and Conduct experiments on 60 cylinders containing simulated TRU waste to determine the effects of matrix depletion on gas generation for transportation. All work for the Gas Generation Matrix Depletion (GGMD) experiment is performed according to the quality objectives established in the test plan and under this Quality Assurance Project Plan (QAPjP)

  20. Plasma influence on throat conductance of the TEXTOR pump limiter ALT-I

    International Nuclear Information System (INIS)

    Hardtke, A.; Finken, K.H.; Reiter, D.; Dippel, K.H.; Goebel, D.M.; McGrath, R.T.; Sagara, A.

    1989-01-01

    On the TEXTOR pump limiter ALT-I conductance measurements for the backstreaming of gas from the pump limiter vessel through the pump limiter entrance have been performed. In these experiments neutral gas has been injected into the pump limiter plenum during a short pulse. The influence of the instreaming plasma results in a reduction of the conductance of the outstreaming gas. For helium the conductance is reduced to about 40% of the molecular conductance when a plasma flux of 0.8 A/cm 2 (T e =T i =11 eV) is streaming into the pump limiter throat. The reduction of the conductance for backstreaming hydrogen and deuterium under the same plasma conditions is smaller; about 70% of the molecular conductance is obtained. This reduction can be explained by an increased recycling of ions which have been produced in the throat back to the neutralizer plate. The experimental results can be reproduced by Monte Carlo neutral transport code calculations if the recycling coefficient is about 0.85 for hydrogen and deuterium and about 0.95 for helium ions. Processes causing these high recycling coefficients are discussed and their influence is estimated. (orig.)

  1. Experimental investigation of thermal-hydraulic performance of PCCS with horizontal tube heat exchangers: single U-tube test

    International Nuclear Information System (INIS)

    Nakamura, Hideo; Anoda, Yoshinari; Arai, Kenji; Kurita, Tomohisa

    2000-01-01

    JAERI and JAPC started a cooperative study to verify performance of a PCCS (Passive Containment Cooling System) using horizontal heat exchanger for next-generation BWR in 1998. A test facility with a horizontal single U-tube was constructed in JAERI in 1999 to investigate fundamental condensation behavior under influences of non-condensable gas. Preliminary pre-test analyses were performed using RELAP5/ MOD3.2.1.2 code to expect the experimental outcomes by incorporating a correlation for condensation degradation because of non-condensable gas by Ueno et al. for better prediction. Preliminary results from both experiments (shakedown) and pre-test analyses indicated that the PCCS using horizontal U-tube heat exchanger is promising. Steam generated under assumed severe accident conditions; steam generation rate approx. = 1% core power, non-condensable gas concentration of 1% and simulated containment vessel pressure of 0.7 MPa, was totally condensed with a small differential pressure across inlet and outlet plenum. Experimental data will be accumulated to develop models and correlations for a better prediction of responses of the PCCS using horizontal heat exchanger during postulated severe accidents. (author)

  2. Operating experience with gas-bearing circulators in a high-pressure helium loop

    International Nuclear Information System (INIS)

    Sanders, J.P.; Gat, U.; Young, H.C.

    1988-01-01

    A high-pressure engineering test loop has been designed and constructed at the Oak Ridge National Laboratory for circulating helium through a test chamber at temperatures to 1,000 deg. C. The purpose of this loop is to determine the thermal and structural performance of proposed components for the primary loops of gas-cooled nuclear reactors. Three gas-bearing circulators, mounted in series, provide a maximum volumetric flow of 0.47 m 3 /s and a maximum head of 78 kJ/kg at operating pressures from 0.1 to 10.7 MPa. Control of gaseous impurities in the circulating gas was the significant operating requirement that dictated the choice of a circulator that is lubricated by the circulating gas. The motor for each circulator is contained within the pressure boundary, and it is cooled by circulating the gas in the motor cavity over water-cooled coils. Each motor is rated at 200 kW at a speed of 23,500 rpm. The circulators have been operated in the loop for more than 5,000 h. The flow of the gas in the loop is controlled by varying the speed of the circulators through the use of individual 250-kVA, solid state power supplies that can be continuously varied in frequency from 50 to 400 Hz. To prevent excessive wear on the gas bearings during startup, the circulator motor accelerates the rotor to 3,000 rpm in less than one second. During operation, no problems associated with the gas bearings, per se, were encountered; however, related problems pointed to design considerations that should be included in future applications of circulators of this type. The primary test that has been conducted in this loop required sustained operation for several weeks without interruption. After a number of unscheduled interruptions, the operating goals were attained. During part of this period, the loop was operated with only two circulators installed in the pressure vessels with a guard installed in the third vessel to protect the closure flange from the gas temperatures. Unattended

  3. Development of a Time Projection Chamber using CF4 gas for relativistic heavy ion experiments

    International Nuclear Information System (INIS)

    Isobe, T.; Hamagaki, H.; Ozawa, K.; Inuzuka, M.; Sakaguchi, T.; Matsumoto, T.; Kametani, S.; Kajihara, F.; Gunji, T.; Kurihara, N.; Oda, S.X.; Yamaguchi, Y.L.

    2006-01-01

    A prototype Time Projection Chamber (TPC) using pure CF 4 gas was developed for possible use in heavy ion experiments. Basic characteristics such as gain, drift velocity, longitudinal diffusion and attenuation length of produced electrons were measured with the TPC. At an electric field of 900V/cm, the drift velocity and longitudinal diffusion for 1cm drift were obtained as 10cm/μs and 60μm, respectively. The relatively large gain fluctuation is explained to be due to the electron attachment process in CF 4 . These characteristics are encouraging for the measurement of the charged particle trajectories under high multiplicity conditions at RHIC

  4. In-silico experiments on characteristic time scale at a shear-free gas-liquid interface in fully developed turbulence

    International Nuclear Information System (INIS)

    Nagaosa, Ryuichi; Handler, Robert A

    2011-01-01

    The purpose of this study is to model scalar transfer mechanisms in a fully developed turbulence for accurate predictions of the turbulent scalar flux across a shear-free gas-liquid interface. The concept of the surface-renewal approximation (Dankwerts, 1951) is introduced in this study to establish the predictive models for the interfacial scalar flux. Turbulent flow realizations obtained by a direct numerical simulation technique are employed to prepare details of three-dimensional information on turbulence in the region very close to the interface. Two characteristic time scales at the interface have been examined for exact prediction of the scalar transfer flux. One is the time scale which is reciprocal of the root-mean-square surface divergence, T γ = (γγ) −1/2 , where γ is the surface divergence. The other time scale to be examined is T S = Λ/V, where Λ is the zero-correlation length of the surface divergence as the interfacial length scale, and V is the root-mean-square velocity fluctuation in the streamwise direction as the interfacial velocity scale. The results of this study suggests that T γ is slightly unsatisfactory to correlate the turbulent scalar flux at the gas-liquid interface based on the surface-renewal approximation. It is also found that the proportionality constant appear to be 0.19, which is different with that observed in the laboratory experiments, 0.34 (Komori, Murakami, and Ueda, 1989). It is concluded that the time scale, T γ , is considered a different kind of the time scale observed in the laboratory experiments. On the other hand, the present in-silico experiments indicate that T s predicts the turbulent scalar flux based on the surface-renewal approximation in a satisfactory manner. It is also elucidated that the proportionality constant for T s is approximately 0.36, which is very close to that found by the laboratory experiments. This fact shows that the time scale T s appears to be essentially the same as the time scale

  5. In-silico experiments on characteristic time scale at a shear-free gas-liquid interface in fully developed turbulence

    Energy Technology Data Exchange (ETDEWEB)

    Nagaosa, Ryuichi [Research Center for Compact Chemical System (CCS), AIST, 4-2-1 Nigatake, Miyagino, Sendai 983-8551 (Japan); Handler, Robert A, E-mail: ryuichi.nagaosa@aist.go.jp [Department of Mechanical Engineering, Texas A and M University, College Station, TX 77843-3123 (United States)

    2011-12-22

    The purpose of this study is to model scalar transfer mechanisms in a fully developed turbulence for accurate predictions of the turbulent scalar flux across a shear-free gas-liquid interface. The concept of the surface-renewal approximation (Dankwerts, 1951) is introduced in this study to establish the predictive models for the interfacial scalar flux. Turbulent flow realizations obtained by a direct numerical simulation technique are employed to prepare details of three-dimensional information on turbulence in the region very close to the interface. Two characteristic time scales at the interface have been examined for exact prediction of the scalar transfer flux. One is the time scale which is reciprocal of the root-mean-square surface divergence, T{sub {gamma}} = ({gamma}{gamma}){sup -1/2}, where {gamma} is the surface divergence. The other time scale to be examined is T{sub S} = {Lambda}/V, where {Lambda} is the zero-correlation length of the surface divergence as the interfacial length scale, and V is the root-mean-square velocity fluctuation in the streamwise direction as the interfacial velocity scale. The results of this study suggests that T{sub {gamma}} is slightly unsatisfactory to correlate the turbulent scalar flux at the gas-liquid interface based on the surface-renewal approximation. It is also found that the proportionality constant appear to be 0.19, which is different with that observed in the laboratory experiments, 0.34 (Komori, Murakami, and Ueda, 1989). It is concluded that the time scale, T{sub {gamma}}, is considered a different kind of the time scale observed in the laboratory experiments. On the other hand, the present in-silico experiments indicate that T{sub s} predicts the turbulent scalar flux based on the surface-renewal approximation in a satisfactory manner. It is also elucidated that the proportionality constant for T{sub s} is approximately 0.36, which is very close to that found by the laboratory experiments. This fact shows

  6. Dynamic modeling of gas turbines in integrated gasification fuel cell systems

    Science.gov (United States)

    Maclay, James Davenport

    2009-12-01

    Solid oxide fuel cell-gas turbine (SOFC-GT) hybrid systems for use in integrated gasification fuel cell (IGFC) systems operating on coal will stretch existing fossil fuel reserves, generate power with less environmental impact, while having a cost of electricity advantage over most competing technologies. However, the dynamic performance of a SOFC-GT in IGFC applications has not been previously studied in detail. Of particular importance is how the turbo-machinery will be designed, controlled and operated in such applications; this is the focus of the current work. Perturbation and dynamic response analyses using numerical SimulinkRTM models indicate that compressor surge is the predominant concern for safe dynamic turbo-machinery operation while shaft over-speed and excessive turbine inlet temperatures are secondary concerns. Fuel cell temperature gradients and anode-cathode differential pressures were found to be the greatest concerns for safe dynamic fuel cell operation. Two control strategies were compared, that of constant gas turbine shaft speed and constant fuel cell temperature, utilizing a variable speed gas turbine. Neither control strategy could eliminate all vulnerabilities during dynamic operation. Constant fuel cell temperature control ensures safe fuel cell operation, while constant speed control does not. However, compressor surge is more likely with constant fuel cell temperature control than with constant speed control. Design strategies that provide greater surge margin while utilizing constant fuel cell temperature control include increasing turbine design mass flow and decreasing turbine design inlet pressure, increasing compressor design pressure ratio and decreasing compressor design mass flow, decreasing plenum volume, decreasing shaft moment of inertia, decreasing fuel cell pressure drop, maintaining constant compressor inlet air temperature. However, these strategies in some cases incur an efficiency penalty. A broad comparison of cycles

  7. Free Surface Water Tunnel (FSWT)

    Data.gov (United States)

    Federal Laboratory Consortium — Description: The Free Surface Water Tunnel consists of the intake plenum, the test section and the exit plenum. The intake plenum starts with a perforated pipe that...

  8. Local content: worldwide trends and the Brazilian experience in the oil and gas sector; Conteudo local: tendencias mundiais e a experiencia brasileira no setor de oleo e gas

    Energy Technology Data Exchange (ETDEWEB)

    Pedrosa Junior, Oswaldo A.; Guimaraes, Paulo Buarque [Associacao Brasileira dos Produtores Independentes de Petroleo e Gas - ABPIP, Rio de Janeiro, RJ (Brazil); Fernandez y Fernandez, Eloi [Organizacao Nacional da Industria do Petroleo, Rio de Janeiro, RJ (Brazil)

    2008-07-01

    In recent years a trend on increasing requirements for local investments has been observed worldwide in the petroleum industry. Host countries expect to have increasing social and economical benefits from the development of the oil and gas industry. This expectation drives at a more comprehensive concept of local content to include commitment with social, industrial, and technological development. The Brazilian experience has shown a lot of emphasis on local industry development. Initiatives from governmental authorities and the private sector have been implemented to increase the local industry participation in the oil and gas projects. The current regulation focus on the full and fair opportunities for the local suppliers and the local content commitment established in the E and P concession agreements. A key issue on promoting local content initiatives is to assure that the competitiveness of the indigenous industry will be developed and preserved. The constraints on building up the local industry competitiveness will be addressed, focusing on the taxation overburden, lack of adequate local financing, and internal structural aspects affecting industrial productivity. In addition to this, the experiences on measuring local content for offshore construction and drilling are highlighted. Technology development and technical capability have been addressed by incentive programs for the O and G sector. Finally, the technology learning process and the regulatory requirements to invest in R and D programs conducted by Brazilian technological institutions are discussed. (author)

  9. Subsurface Noble Gas Sampling Manual

    Energy Technology Data Exchange (ETDEWEB)

    Carrigan, C. R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sun, Y. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-09-18

    The intent of this document is to provide information about best available approaches for performing subsurface soil gas sampling during an On Site Inspection or OSI. This information is based on field sampling experiments, computer simulations and data from the NA-22 Noble Gas Signature Experiment Test Bed at the Nevada Nuclear Security Site (NNSS). The approaches should optimize the gas concentration from the subsurface cavity or chimney regime while simultaneously minimizing the potential for atmospheric radioxenon and near-surface Argon-37 contamination. Where possible, we quantitatively assess differences in sampling practices for the same sets of environmental conditions. We recognize that all sampling scenarios cannot be addressed. However, if this document helps to inform the intuition of the reader about addressing the challenges resulting from the inevitable deviations from the scenario assumed here, it will have achieved its goal.

  10. Database on gas migration tests through bentonite buffer material

    International Nuclear Information System (INIS)

    Tanai, Kenji

    2009-02-01

    Carbon steel is a candidate material for an overpack for geological disposal of high-level radioactive waste in Japan. The corrosion of the carbon steel overpack in aqueous solution under anoxic conditions will cause the generation of hydrogen gas, which may affect hydrological and mechanical properties of the bentonite buffer. To evaluate such an effect of gas generation, it is necessary to develop a model of gas migration through bentonite buffer material taking account of data obtained from experiments. The gas migration experiments under both unsaturated and saturated conditions have been carried out to clarify the fundamental characteristics of bentonite for gas migration. This report compiles the experimental data obtained from gas migration tests for buffer material which has been conducted by JAEA until December, 2007. A CD-ROM is attached as an appendix. (author)

  11. Premixed direct injection disk

    Science.gov (United States)

    York, William David; Ziminsky, Willy Steve; Johnson, Thomas Edward; Lacy, Benjamin; Zuo, Baifang; Uhm, Jong Ho

    2013-04-23

    A fuel/air mixing disk for use in a fuel/air mixing combustor assembly is provided. The disk includes a first face, a second face, and at least one fuel plenum disposed therebetween. A plurality of fuel/air mixing tubes extend through the pre-mixing disk, each mixing tube including an outer tube wall extending axially along a tube axis and in fluid communication with the at least one fuel plenum. At least a portion of the plurality of fuel/air mixing tubes further includes at least one fuel injection hole have a fuel injection hole diameter extending through said outer tube wall, the fuel injection hole having an injection angle relative to the tube axis. The invention provides good fuel air mixing with low combustion generated NOx and low flow pressure loss translating to a high gas turbine efficiency, that is durable, and resistant to flame holding and flash back.

  12. Simulation experiments on the radial pool growth in gas-releasing melting system

    International Nuclear Information System (INIS)

    Farhadieh, R.; Purviance, R.; Carlson, N.

    1983-01-01

    Following an HCDA, molten core-debris can contact the concrete foundation of the reactor building resulting in a molten UO 2 /concrete interaction and considerable gas release. The released gas can pressurize the containment building potentially leading to radiological releases. Furthermore, directional growth of the molten core-debris pool can reduce the reactor building structural integrity. To implement design changes that insure structural integrity, an understanding of the thermal-hydraulic and mass-transfer process associated with such a growth is most desirable. Owing to the complex nature of the combined heat, mass, and hydrodynamic processes associated with the two-dimensional problem of gas release and melting, the downward and radial penetration problems have been investigated separately. The present experimental study addresses the question of sideward penetration of the molten core debris into a gas-releasing, meltable, miscible solid

  13. Measured gas and particle temperatures in VTT's entrained flow reactor

    DEFF Research Database (Denmark)

    Clausen, Sønnik; Sørensen, L.H.

    2006-01-01

    Particle and gas temperature measurements were carried out in experiments on VTTs entrained flow reactor with 5% and 10% oxygen using Fourier transform infrared emission spectroscopy (FTIR). Particle temperature measurements were performed on polish coal,bark, wood, straw particles, and bark...... and wood particles treated with additive. A two-color technique with subtraction of the background light was used to estimate particle temperatures during experiments. A transmission-emission technique was used tomeasure the gas temperature in the reactor tube. Gas temperature measurements were in good...... agreement with thermocouple readings. Gas lines and bands from CO, CO2 and H2O can be observed in the spectra. CO was only observed at the first measuring port (100ms) with the strongest CO-signal seen during experiments with straw particles. Variations in gas concentration (CO2 and H2O) and the signal from...

  14. 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.

  15. A study on self-excited sloshing due to the fluid discharge over a flexible weir

    International Nuclear Information System (INIS)

    Nagakura, Hiroshi; Kaneko, Shigehiko.

    1995-01-01

    An analytical model for the fluid-elastic instability as observed in Super-Phenix-1 LMFBR is proposed. This fluid-structure system is constituted by the flexible weir and adjoining fluid plenums, and the fluid is discharged from the upstream plenum to the downstream plenum over a flexible weir. The characteristic equation of the system is derived for the case in which the weir vibrates at the frequency of the downstream plenum sloshing. The effects of the fluid level difference between the upstream and the downstream plenum and weir rigidity are examined, and the mechanism for instability is discussed. (author)

  16. CHARACTERISTICS OF SELF-LEVELING BEHAVIOR OF DEBRIS BEDS IN A SERIES OF EXPERIMENTS

    Directory of Open Access Journals (Sweden)

    SONGBAI CHENG

    2013-06-01

    Full Text Available During a hypothetical core-disruptive accident (CDA in a sodium-cooled fast reactor (SFR, degraded core materials can form roughly conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of the core material pool. However, coolant boiling may ultimately lead to leveling of the debris bed, which is crucial to the relocation of the molten core and heat-removal capability of the debris bed. To clarify the mechanisms underlying this self-leveling behavior, a large number of experiments were performed within a variety of conditions in recent years, under the constructive collaboration between the Japan Atomic Energy Agency (JAEA and Kyushu University (Japan. The present contribution synthesizes and gives detailed comparative analyses of those exper