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Sample records for reactor outlet plenums

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

    One aspect of the AP1000{sup Registered-Sign} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is the reduction in the number of reactor vessel outlet nozzles/hot legs leaving the upper plenum from three to two. With regard to fuel performance, this design difference creates a different flow field in the AP1000 reactor vessel upper plenum (the region above the core). The flow exiting core and entering the upper plenum must turn 90 Degree-Sign , flow laterally through the upper plenum around support structures, and exit through one of the two outlet nozzles. While the flow in the top of the core is mostly axial, there is some lateral flow component as the core flow reacts to the flow field and pressure distribution in the upper plenum. The pressure distribution in the upper plenum varies laterally depending upon various factors including the proximity to the outlet nozzles. To determine how the lateral flow in the top of the AP1000 core compares to current Westinghouse reactors, a computational fluid dynamics (CFD) model of the flow in the upper portion of the AP1000 reactor vessel including the top region of the core, the upper plenum, the reactor vessel outlet nozzles, and a portion of the hot legs was created. Due to geometric symmetry, the computational domain was reduced to a quarter (from the top view) that includes Vulgar-Fraction-One-Quarter of the top of the core, Vulgar-Fraction-One-Quarter of the upper plenum, and Vulgar-Fraction-One-Half of an outlet nozzle. Results from this model include predicted velocity fields and pressure distributions throughout the model domain. The flow patterns inside and around guide tubes clearly demonstrate the influence of lateral flow due to the presence of the outlet nozzles. From these results, comparisons of AP1000 flow versus current Westinghouse plants were performed. Field performance

  17. 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

  18. 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

  19. 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.

  20. 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

  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. 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

  4. 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

  5. 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

  6. 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

  7. 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.)

  8. 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)

  9. 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)

  10. Flow model study of 'Monju' reactor vessel

    International Nuclear Information System (INIS)

    Miyaguchi, Kimihide

    1980-01-01

    In the case of designing the structures in nuclear reactors, various problems to be considered regarding thermo-hydrodynamics exist, such as the distribution of flow quantity and the pressure loss in reactors and the thermal shock to inlet and outlet nozzles. In order to grasp the flow characteristics of coolant in reactors, the 1/2 scale model of the reactor structure of ''Monju'' was attached to the water flow testing facility in the Oarai Engineering Center, and the simulation experiment has been carried out. The flow characteristics in reactors clarified by experiment and analysis so far are the distribution of flow quantity between high and low pressure regions in reactors, the distribution of flow quantity among flow zones in respective regions of high and low pressure, the pressure loss in respective parts in reactors, the flow pattern and the mixing effect of coolant in upper and lower plenums, the effect of the twisting angle of inlet nozzles on the flow characteristics in lower plenums, the effect of internal cylinders on the flow characteristics in upper plenums and so on. On the basis of these test results, the improvement of the design of structures in reactors was made, and the confirmation test on the improved structures was carried out. The testing method, the calculation method, the test results and the reflection to the design of actual machines are described. (Kako, I.)

  11. Laboratory determination of normal operating flow rates with enlarged outlet fittings -- BDF reactors

    Energy Technology Data Exchange (ETDEWEB)

    Waters, E.D.

    1960-02-02

    Experiments have been conducted in the Hydraulics Laboratory, at the request of IPD`s Mechanical Development-A Operation, to determine the energy losses of various enlarged outlet fitting combinations. These experiments were conducted an steady state runs and allow the determination of the normal operating point (flow rate) of a reactor process channel under selected conditions of front header pressure and fuel charge. No attempt is made to make a mechanical or economic evaluation of the particular fitting combinations, although observations were noted which might bear on this evaluation. It is very important for the reader to bear in mind that changing outlet fittings will definitely affect the reactor tube power limits and outlet vater temperature limits. The size of the outlet fittings largely determines the present outlet temperature limits of the old reactors. The flow characteristics of these present fittings cause some degree of pressurization to suppress boiling on the fuel charge and also cause dual Panellit trip protection for certain flow changes and for power surges. Enlargement of the outlet fittings may actually reduce the allowable outlet coolant temperature limits. Since these effects cannot be determined on the apparatus used in these experiments, a complete discussion of this point is not included in this report. However, the seriousness of these effects should be known and carefully analyzed before a final selection of enlarged outlet fittings in made. This report will be one of a series. New reports in the series will be issued as data are obtained for other such outlet fitting combinations or for new concepts of outlet fitting assemblies such as the new nozzle being developed by C. E. Trantz for use on F-reactor stuck gunbarrel tubes.

  12. 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.

  13. 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.

  14. Periodic large-amplitude thermal oscillations occurring in a buoyant plume

    International Nuclear Information System (INIS)

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

    1983-01-01

    Reactor events such as N-1 loop operation in conjunction with a leaky check valve in the down loop can cause flow to be convected back into the reactor outlet nozzle/piping region and to be back-flushed into the reactor outlet plenum. The preceding results in a temperature difference between pipe inflow and plenum. This temperature difference causes buoyancy forces which if large enough can cause: a pipe backflow and recirculation loop; and a thermal plume in the plenum. Both phenomena are being studied because they can produce undesirable pipe, nozzle and plenum wall thermal distributions, and hence undesirable thermal stresses. This paper discusses some features of the plume

  15. 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)

  16. 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

  17. 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

  18. 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)

  19. Numerical analysis of temperature fluctuation in core outlet region of China experimental fast reactor

    International Nuclear Information System (INIS)

    Zhu Huanjun; Xu Yijun

    2014-01-01

    The temperature fluctuation in core outlet region of China Experimental Fast Reactor (CEFR) was numerically simulated by the CFD software Star CCM+. With the core outlet temperatures, flows etc. under rated conditions given as boundary conditions, a 1/4 region model of the reactor core outlet region was established and calculated using LES method for this problem. The analysis results show that while CEFR operates under rated conditions, the temperature fluctuation in lower part of core outlet region is mainly concentrated in area over the edge components (steel components, control rod assembly), and one in upper part is remarkable in area above all the components. The largest fluctuation amplitude is 19 K and the remarkable frequency is below 5 Hz, and it belongs to typically low frequency fluctuation. The conclusion is useful for further experimental work. (authors)

  20. 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

  1. 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

  2. Experiment study on thermal mixing performance of HTR-PM reactor outlet

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Yangping, E-mail: zhouyp@mail.tsinghua.edu.cn [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); Hao, Pengfei [School of Aerospace, Tsinghua University, Beijing 100084 (China); Li, Fu; Shi, Lei [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China); He, Feng [School of Aerospace, Tsinghua University, Beijing 100084 (China); Dong, Yujie; Zhang, Zuoyi [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, the Key Laboratory of Advanced Reactor Engineering and Safety, Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2016-09-15

    A model experiment is proposed to investigate the thermal mixing performance of HTR-PM reactor outlet. The design of 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 mixing structure, hot gas duct, exhaust pipe system and I&C system. 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 results show the mixing efficiency of the test facility is higher than that required by the steam generator of HTR-PM, which indicates that the thermal mixing structure of HTR-PM fulfills its design requirement.

  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. 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

  5. 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.)

  6. Simulation of inlet and outlet riser break sequences in the N Reactor

    International Nuclear Information System (INIS)

    Bolander, M.A.; Fletcher, C.D.

    1988-02-01

    This report documents work performed at the Idaho National Engineering Laboratory (INEL) in support of the Westinghouse Hanford Company safety analyses of the N Reactor. The RELAP5/MOD2 computer code was used in analyzing two hypothetical transients. The computer code was modified specifically to simulate the refill behavior in the N Reactor process tubes. The transients analyzed were a double-ended rupture of an inlet riser column and a double-ended rupture of an outlet riser column

  7. 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.

  8. 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)

  9. 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)

  10. 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)

  11. 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.

  12. Numerical Simulation of Two-branch Hot Gas Mixing at Reactor Outlet of HTR-PM

    International Nuclear Information System (INIS)

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

    2014-01-01

    A series of two-branch model experiment has been finished to investigate the thermal mixing efficiency of the HTR-PM reactor outlet. This paper introduces the numerical simulation on the design of thermal mixing structure of HTR-PM and the test facility with Fluent software. The profiles of temperature, pressure and velocity in the mixing structure design and the test facility are discussed by comparing with the model experiment results. The numerical simulation results of the test facility have good agreement to the experiment results. In addition, the thermal-fluid characters obtained by numerical simulation show the thermal mixing structure of HTR-PM has similarity with the test facility. Finally, it is concluded that the thermal mixing design at HTR-PM reactor outlet can fulfilled the requirements for high thermal mixing efficiency and appropriate pressure drop. (author)

  13. 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)

  14. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    Scott, D.

    1979-01-01

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  15. 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

  16. Scale-model characterization of flow-induced vibrational response of FFTF reactor internals

    International Nuclear Information System (INIS)

    Ryan, J.A.; Mahoney, J.J.

    1980-10-01

    Fast Test Reactor core internal and peripheral components were assessed for flow-induced vibrational characteristics under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup as an integral part of the Fast Test Reactor Vibration Program. The Hydraulic Core Mockup was an 0.285 geometric scale model of the Fast Test Reactor internals designed to simulate prototype vibrational and hydraulic characteristics. Using water to simulate sodium coolant, vibrational characteristics were measured and determined for selected model components over the scaled flow range of 36 to 110%. Additionally, in-situ shaker tests were conducted on selected Hydraulic Core Mockup outlet plenum components to establish modal characteristics. Most components exhibited resonant response at all test flow rates; however, the measured dynamic response was neither abnormal nor anomalously flow-rate dependent, and the predicted prototype components' response were deemed acceptable

  17. 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.)

  18. 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)

  19. 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)

  20. 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

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

    International Nuclear Information System (INIS)

    Paul, D.D.

    1976-01-01

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

  2. 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

  3. Effect of cross-flow direction of coolant on film cooling effectiveness with one inlet and double outlet hole injection

    Directory of Open Access Journals (Sweden)

    Guangchao Li

    2012-12-01

    Full Text Available In order to study the effect of cross-flow directions of an internal coolant on film cooling performance, the discharge coefficients and film cooling effectiveness with one inlet and double outlet hole injections were simulated. The numerical results show that two different cross-flow directions of the coolant cause the same decrease in the discharge coefficients as that in the case of supplying coolant by a plenum. The different proportion of the mass flow out of the two outlets of the film hole results in different values of the film cooling effectiveness for three different cases of coolant supplies. The film cooling effectiveness is the highest for the case of supplying coolant by the plenum. At a lower blowing ratio of 1.0, the film cooling effectiveness with coolant injection from the right entrance of the passage is higher than that from the left entrance of the passage. At a higher blowing ratio of 2.0, the opposite result is found.

  4. 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

  5. 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)

  6. 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)

  7. 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)

  8. 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.)

  9. 3-D thermal hydraulic analysis of transient heat removal from fast reactor core using immersion coolers

    International Nuclear Information System (INIS)

    Chvetsov, I.; Volkov, A.

    2000-01-01

    For advanced fast reactors (EFR, BN-600M, BN-1600, CEFR) the special complementary loop is envisaged in order to ensure the decay heat removal from the core in the case of LOF accidents. This complementary loop includes immersion coolers that are located in the hot reactor plenum. To analyze the transient process in the reactor when immersion coolers come into operation one needs to involve 3-D thermal hydraulics code. Furthermore sometimes the problem becomes more complicated due to necessity of simulation of the thermal hydraulics processes into the core interwrapper space. For example on BN-600M and CEFR reactors it is supposed to ensure the effective removal of decay heat from core subassemblies by specially arranged internal circulation circuit: 'inter-wrapper space'. For thermal hydraulics analysis of the transients in the core and in the whole reactor including hot plenum with immersion coolers and considering heat and mass exchange between the main sodium flow and sodium that moves in the inter-wrapper space the code GRIFIC (the version of GRIF code family) was developed in IPPE. GRIFIC code was tested on experimental data obtained on RAMONA rig under conditions simulating decay heat removal of a reactor with the use of immersion coolers. Comparison has been made of calculated and experimental result, such as integral characteristics (flow rate through the core and water temperature at the core inlet and outlet) and the local temperatures (at thermocouple location) as well. In order to show the capabilities of the code some results of the transient analysis of heat removal from the core of BN-600M - type reactor under loss-of-flow accident are presented. (author)

  10. Thermal stratification of sodium in the BN 600 reactor

    International Nuclear Information System (INIS)

    Obmelukhin, J.A.; Obukhov, P.I.; Rinejskij, A.A.; Sobolev, V.A.; Sherbakov, S.I.

    1983-01-01

    The signs of thermal stratification of sodium in the BN 600 reactor upper plenum revealed by the analysis of standard temperature sensors' readings are defined. The initial conditions for existence of different temperature sodium layers are given. Two approaches for realizing on a computer of equations describing sodium motion in the upper plenum of the reactor are presented. (author)

  11. Analysis of systematic error deviation of water temperature measurement at the fuel channel outlet of the reactor Maria

    International Nuclear Information System (INIS)

    Bykowski, W.

    2000-01-01

    The reactor Maria has two primary cooling circuits; fuel channels cooling circuit and reactor pool cooling circuit. Fuel elements are placed inside the fuel channels which are parallely linked in parallel, between the collectors. In the course of reactor operation the following measurements are performed: continuous measurement of water temperature at the fuel channels inlet, continuous measurement of water temperature at the outlet of each fuel channel and continuous measurement of water flow rate through each fuel channel. Based on those thermal-hydraulic parameters the instantaneous thermal power generated in each fuel channel is determined and by use of that value the thermal balance and the degree of fuel burnup is assessed. The work contains an analysis concerning estimate of the systematic error of temperature measurement at outlet of each fuel channel and so the erroneous assessment of thermal power extracted in each fuel channel and the burnup degree for the individual fuel element. The results of measurements of separate factors of deviations for the fuel channels are enclosed. (author)

  12. Enhancing load-following and/or spectral shift capability in single-sparger natural circulation boiling water reactors

    International Nuclear Information System (INIS)

    Oosterkamp, W.J.

    1992-01-01

    This patent describes a method for obtaining load-following capability in a coiling water reactor (BWR) wherein housed within a reactor pressure vessel (RPV) is a nuclear core disposed within a shroud having a shroud head and which with the RPV defines an annulus region disposed beneath the nuclear core, an upper steam dome connected to a steam outlet in the RPV, a core upper plenum formed within the shroud head and disposed atop the nuclear core, a chimney mounted atop the shroud head and in fluid communication with the core upper plenum and with a steam separator having a skirt which is in fluid communication with the steam dome, the region outside of the chimney defining a downcomer region, there being a water level established therein under normal operation of the BWR, and the RPV containing a feedwater inlet. It comprises: disposing a single sparger connected to the feedwater inlet above the steam separator skirt bottom about the interior circumference of the RPV at an elevation at approximately the water level established during normal operation of the BWR; and adjusting the feedwater flow through the inlet and into the sparger to vary the water level to be above, at or below the elevational location of the sparger in response to load-following need

  13. Method for determining the outlet temperature of fuel assemblies unsupplied with thermometer in WWER-440 reactors

    International Nuclear Information System (INIS)

    Miko, S.; Kalya, Z.; Hamvas, I.

    1987-09-01

    The paper outlines a method for the evaluation of the outlet temperatures of fuel assemblies unsupplied with thermometer in WWER-440 reactors. The process is based on interpolation of directly measured assembly temperatures. A quantitative comparison of the errors of described algorithm to those of standard plant-computer interpolation rutine is also presented. (author)

  14. Study on applicability of PIV measurement to natural convection in a scaled reactor vessel model

    International Nuclear Information System (INIS)

    Murakami, Takahiro; Koga, Tomonari; Eguchi, Yuzuru; Watanabe, Osamu

    2009-01-01

    The applicability of Particle Image Velocimetry (PIV) to natural convection in the plenum of a scaled water test model of the Japan Sodium-cooled Fast Reactor (JSFR) is studied in the paper. PIV measurement of such a buoyancy-driven flow in a geometrically complicated vessel is difficult in general, because the detection rate of tracer particles tends to decrease, and the noisy optical reflection to increase. In our measurements, tracer particles are adequately seeded in the hot plenum and particle images are captured by using a double-pulsed Nd:YAG laser and a high-speed camera. Then, image-processing techniques are employed to eliminate unphysical velocity vectors and unnecessary background images. The PIV results have shown that clear flow pattern can be extracted by time-averaging 300 sets of instantaneous PIV data in spite of highly fluctuating features of velocity in space and time. Moreover, the evaluation of the statistical quantities such as variance, skewness, and kurtosis has revealed the characteristic of the non-stationary spouting flows at the heater outlet. (author)

  15. Thermal-hydraulic modeling of porous bed reactors

    International Nuclear Information System (INIS)

    Araj, K.J.; Nourbakhsh, H.P.

    1987-01-01

    Optimum design of nuclear reactor cores requires an iterative approach between the thermal-hydraulic, neutronic, and operational analysis. This paper will concentrate on the thermal-hydraulic behavior of a hydrogen-cooled small particle bed reactor (PBR). The PBR core modeled here consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flows, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit to a common plenum. A fast running one-dimensional lumped-parameter steady-state code (FTHP) was developed to evaluate the effects of design changes in fuel assembly and power distribution. Another objective for the code was to investigate various methods of coolant control to minimize hot channel effects and maximize outlet temperatures

  16. 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)

  17. Loop-type FBR reactor

    International Nuclear Information System (INIS)

    Ogura, Kenji; Kimura, Kimitaka; Jinbo, Masaichi; Hirayama, Hiroshi; Taguchi, Junzo; Hirata, Noriaki; Ozaki, Kenji; Maruyama, Shigeki.

    1996-01-01

    The inside of a vessel of an intermediate heat exchanger is divided vertically by a partition wall into a high temperature plenum region and a low temperature plenum region, a perforated horizontal plate is disposed in a horizontal direction at the upper portion and a flow shroud is disposed so as to surround the upper outside of the intermediate heat exchanger while passing through a lid from a perforated hole of the perforated horizontal plate. In addition, there is disposed a cylinder passing through the partition wall and the horizontal perforated plate for inserting a liquid surface penetrating equipment. The cylinder has an upper end opened above the liquid level of a liquid metal during normal operation and below the liquid level of the liquid metal during shut down of the reactor, and the lower end is opened in a lower plenum region. Vibrations of liquid level due to the high temperature liquid metal inflown from a hot leg pipeline to the inside of the vessel of the intermediate heat exchanger are suppressed by the perforated horizontal plate during reactor operation. On the other hand, upon shut down of the reactor, since the liquid level rises up to the upper portion of the cylinder, the liquid metal at low temperature inflows into the lower plenum region, and the liquid metal at high temperature above the horizontal perforated plate is eliminated in an early stage. (N.H.)

  18. FBR type reactors

    International Nuclear Information System (INIS)

    Maemoto, Junko.

    1985-01-01

    Purpose: To moderate abrupt temperature change near the inner walls of a suspended body thereby prevent thermal shocks and thermal deformations to structural materials. Constitution: High temperature coolants during ordinary operation of the nuclear reactor flow from the reactor core through the flow holes of the suspended body and from the upper plenum into an intermediate heat exchanger. The temperature of the coolants is lowered with heat exchanging effect with secondary coolants in the heat exchange and the coolants are then flow through the lower plenum into the reactor core and heated again. Upon generation of reactor scram, the temperature of the coolants at the exit of the reactor core is reduced abruptly and the flow rate is lowered due to the pump coast down. However, mixing of the coolants in the suspended body is accelerated by the coolants at high temperature flowing out of the flow holes and the coolants at the low temperature flowing from the flow hole group, to reduce the temperature difference and moderate the stratification flow forming an abrupt temperature slope. (Yoshihara, H.)

  19. Molten core material holding device in a nuclear reactor

    International Nuclear Information System (INIS)

    Nakamura, Hisashi; Tanaka, Nobuo; Takahashi, Katsuro.

    1985-01-01

    Purpose: To improve the function of cooling to hold molten core materials in a molten core material holding device. Constitution: Plenum structures are formed into a pan-like configuration, in which liners made of metal having high melting point and relatively high heat conductivity such as tantalum, tungsten, rhenium or alloys thereof are integrally appended to hold and directly cool the molten reactor core materials. Further, a plurality of heat pipes, passing through the plenum structures, facing the cooling portion thereof to the coolants at the outer side and immersing the heating portion into the molten core materials fallen to deposit in the inner liners are disposed radially. Furthermore, heat pipes embodded in the plenum structure are disposed in the same manner below the liners. Thus, the plenum structures and the molten reactor core materials can be cooled at a high efficiency. (Seki, T.)

  20. 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.

  1. Simulation of In-Vessel Corium Retention through External Reactor Vessel Cooling for SMART using SIMPLE

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jin-Sung; Son, Donggun; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Thermal load analysis from the corium pool to the outer reactor vessel in the lower plenum of the reactor vessel is necessary to evaluate the effect of the IVR-ERVC during a severe accident for SMART. A computational code called SIMPLE (Sever Invessel Melt Progression in Lower plenum Environment) has been developed for analyze transient behavior of molten corium in the lower plenum, interaction between corium and coolant, and heat-up and ablation of reactor vessel wall. In this study, heat load analysis of the reactor vessel for SMART has been conducted using the SIMPLE. Transient behavior of the molten corium in the lower plenum and IVR-ERVC for SMART has been simulated using SIMPLE. Heat flux from the corium pool to the outer reactor vessel is concentrated in metallic layer by the focusing effect. As a result, metallic layer shows higher temperature than the oxidic layer. Also, vessel wall of metallic layer has been ablated by the high in-vessel temperature. Ex-vessel temperature of the metallic layer was maintained 390 K and vessel thickness was maintained 14 cm. It means that the reactor vessel integrity is maintained by the IVR-ERVC.

  2. Contributions to and expectations from the CRP - Argonne National Laboratory (USA)

    International Nuclear Information System (INIS)

    Cahalan, J.E.

    2007-01-01

    For us, the chief benefit of the CRP will be validation of multidimensional fluid dynamics capabilities for analysis of outlet plenum temperature distributions. As reactor designers seek new fuel handling features to reduce costs, upper internal structure configurations are becoming more compact, and higher fidelity analysis techniques are required to assess thermal stresses. Argonne currently has 1) a reactor systems analysis code with an experimentally-based model for plenum stratification, 2) the COMMIX code (parent of the JAEA AQUA code), and 3) commercial fluid dynamics analysis codes. It is anticipated that all or some combination of these capabilities will be employed to perform the CRP analysis

  3. 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.))

  4. 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)

  5. 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.

  6. Evaluation method for core thermohydraulics during natural circulation in fast reactors numerical predictions of inter-wrapper flow

    International Nuclear Information System (INIS)

    Kamide, H.; Kimura, N.; Miyakoshi, H.; Nagasawa, K.

    2001-01-01

    Decay heat removal using natural circulation is one of the important functions for the safety of fast reactors. As a decay heat removal system, direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this design, dumped heat exchanger provides cold sodium and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such natural circulation phenomena in a reactor core has been developed, which models each subassembly as a rectangular duct with gap region between the subassemblies and also the upper plenum in a reactor vessel. This numerical simulation method was verified based on experimental data of a sodium test using 7- subassembly core model and also a water test which simulates IWF using the 1/12 sector model of a reactor core. We applied the estimation method to the natural circulation in a 600 MW class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. It is desired for prediction methods on the natural circulation to simulate these phenomena. (author)

  7. 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.

  8. 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

  9. 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

  10. 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.

  11. Computational simulation of the natural circulation occurring in an experimental test section of a pool type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nascimento, Francisco R.T. do; Lima Junior, Carlos A.S.; Oliveira, Andre F.S. de; Affonso, Renato R.W.; Faccini, Jose L.H.; Moreira, Maria L., E-mail: rogerio.tdn@gmail.com, E-mail: souzalima_ca@ien.gov.br, E-mail: oliveira.afelipe@gmail.com, E-mail: raoniwa@yahoo.com.br, E-mail: faccini@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The present work presents a computational simulation of the natural circulation phenomenon developing in an experimental test section of a pool type research reactor. The test section has been designed using a reduced scale in height 1:4.7 in relation to a pool type 30 MW research reactor prototype. It comprises a cylindrical vessel, which is opened to atmosphere, and representing the reactor pool; a natural circulation pipe, a lower plenum, and a heater containing electrical resistors in rectangular plate format, which represents the fuel elements, with a chimney positioned on the top of the resistor assembly. In the computational simulation, it was used a commercial CFD software, without any turbulence model. Besides, in the presence of the natural circulation, a laminar flow has been assumed and the equations of the mass conservation, momentum and energy were solved by the finite element method. In addition, the results of the simulation are presented in terms of velocities and temperatures differences, respectively: at inlet and outlet of the heater and of the natural circulation pipe. (author)

  12. Computational simulation of the natural circulation occurring in an experimental test section of a pool type research reactor

    International Nuclear Information System (INIS)

    Nascimento, Francisco R.T. do; Lima Junior, Carlos A.S.; Oliveira, Andre F.S. de; Affonso, Renato R.W.; Faccini, Jose L.H.; Moreira, Maria L.

    2015-01-01

    The present work presents a computational simulation of the natural circulation phenomenon developing in an experimental test section of a pool type research reactor. The test section has been designed using a reduced scale in height 1:4.7 in relation to a pool type 30 MW research reactor prototype. It comprises a cylindrical vessel, which is opened to atmosphere, and representing the reactor pool; a natural circulation pipe, a lower plenum, and a heater containing electrical resistors in rectangular plate format, which represents the fuel elements, with a chimney positioned on the top of the resistor assembly. In the computational simulation, it was used a commercial CFD software, without any turbulence model. Besides, in the presence of the natural circulation, a laminar flow has been assumed and the equations of the mass conservation, momentum and energy were solved by the finite element method. In addition, the results of the simulation are presented in terms of velocities and temperatures differences, respectively: at inlet and outlet of the heater and of the natural circulation pipe. (author)

  13. 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)

  14. Savannah River Site reactor hardware design modification study

    International Nuclear Information System (INIS)

    Fisher, J.E.

    1990-01-01

    A study was undertaken to assess the merits of proposed design modifications to the Savannah River Site (SRS) reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. System recovery potential was evaluated for break locations at the pump suction, the pump discharge, and the plenum inlet. The code version used was RELAP5/MOD2.5 version 3d3, a preliminary version of RELAP5/MOD3. The model was a three-dimensional representation of the K-Reactor water plenum and moderator tank. It included explicit representations of all six loops, which were based on the configuration of L-Reactor. A combination of features is recommended to ensure liquid inventory recovery for all break locations. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 7 refs., 10 figs., 2 tabs

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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

  1. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  2. 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

  3. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  4. Computational fluid dynamic analysis of core bypass flow phenomena in a prismatic VHTR

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Johnson, Richard; Schultz, Richard

    2010-01-01

    The core bypass flow in a prismatic very high temperature reactor (VHTR) is an important design consideration and can have considerable impact on the condition of reactor core internals including fuels. The interstitial gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The occurrence of hot spots in the core and lower plenum and hot streaking in the lower plenum (regions of very hot gas flow) are affected by bypass flow. In the present study, three-dimensional computational fluid dynamic (CFD) calculations of a typical prismatic VHTR are conducted to better understand bypass flow phenomena and establish an evaluation method for the reactor core using the commercial CFD code FLUENT. Parametric calculations changing several factors in a one-twelfth sector of a fuel column are performed. The simulations show the impact of each factor on bypass flow and the resulting flow and temperature distributions in the prismatic core. Factors include inter-column gap-width, turbulence model, axial heat generation profile and geometry change from irradiation-induced shrinkage in the graphite block region. It is shown that bypass flow provides a significant cooling effect on the prismatic block and that the maximum fuel and coolant channel outlet temperatures increase with an increase in gap-width, especially when a peak radial factor is applied to the total heat generation rate. Also, the presence of bypass flow causes a large lateral temperature gradient in the block and also dramatically increases the variation in coolant channel outlet temperatures for a given block that may have repercussions on the structural integrity of the graphite, the neutronics and the potential for hot streaking and hot spots occurring in the lower plenum.

  5. Spiral-shaped disinfection reactors

    KAUST Repository

    Ghaffour, NorEddine; Ait-Djoudi, Fariza; Naceur, Wahib Mohamed; Soukane, Sofiane

    2015-01-01

    This disclosure includes disinfection reactors and processes for the disinfection of water. Some disinfection reactors include a body that defines an inlet, an outlet, and a spiral flow path between the inlet and the outlet, in which the body

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

    International Nuclear Information System (INIS)

    Franklin, R.; Rodriguez, C.

    1982-06-01

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

  7. 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)

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

    International Nuclear Information System (INIS)

    Shen Xiuzhong; Yu Pingan; Yang Guanyue

    1996-01-01

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

  9. A preliminary CATHENA thermalhydraulic model of the Canadian SCWR for safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wang, D.F.; Wang, S., E-mail: wangd@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-06-15

    The supercritical water-cooled reactor (SCWR) is one of six reactor concepts under development in the Generation-IV International Forum (GIF). As a member of GIF, Canada is developing a pressure-tube type SCWR, which has the potential to fulfill all major GIF goals on enhanced safety, sustainability, economics, and proliferation resistance. The system thermalhydraulics code CATHENA will be used in the safety analyses for the Canadian SCWR. Based on the current conceptual design of the Canadian SCWR, a CATHENA idealization has been developed. This model includes all 336 fuel channels with a detailed model of heat transfer in the reactor core. Also modeled are the main pumps, inlet plenum, outlet plenum, turbines, and heavy water moderator. In this paper, the CATHENA idealization of the Canadian SCWR conceptual design is described. Simulation results for steady-state normal operations are also presented for the current Canadian SCWR conceptual design. (author)

  10. Flow visualization study of two phase flow in a single bend outlet feeder pipe and horizontal annulus of outlet end-fitting of a CANDU reactor

    International Nuclear Information System (INIS)

    Supa-Amornkul, S.; Lister, D.H.; Steward, F.R.

    2005-01-01

    'Full text:' In CANDU-6 reactors, the pressurized high-temperature coolant flows through 380 fuel channels passing horizontally through the core. Each end of a fuel channel has a stainless steel annular end-fitting connected to a carbon steel feeder pipe. The outlet coolant, which is at 310 o C with up to 0.30 steam voidage, turns through 90 o as it passes from flow in the annular end-fitting to pipe flow in the feeder via a Grayloc connector. Since 1996, several CANDU stations have reported excessive corrosion of their outlet feeder pipes; especially between the first metre, which consisted of single or double bends. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow accelerated corrosion. In order to understand the hydrodynamics of the coolant in the outlet feeders by flow-visualization, a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream annular end-fitting were fabricated. The feeder consisted of a 54 mm inside diameter acrylic pipe with a 73 o bend, connecting to an acrylic simulation of a Grayloc flanged fitting and annular end-fitting. The annular end-fitting consisted of an inner pipe, 110 mm outer diameter, and an outer pipe, 150 mm inner diameter, both 190.7 cm long in length. The tests were performed with water and air at atmospheric pressure and room temperature. The maximum water volumetric flow rate was 19 L/s and the volume fraction of air varied from 0.05 to 0.56. The phase distributions within the feeder pipe and along the length of the annulus were investigated with a digital video recorder. Size, concentration and velocity of the air bubbles at particular locations were studied with a high-speed digital still camera and a high-speed digital video camera. Phase distributions and variations in bubble size with velocity were determined. Particular attention was paid to the flow pattern at the inside of the bend, where a CFD

  11. Flow visualization study of two phase flow in a single bend outlet feeder pipe and horizontal annulus of outlet end-fitting of a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Supa-Amornkul, S.; Lister, D.H.; Steward, F.R. [Univ. of New Brunswick, Fredericton, New Brunswick (Canada)]. E-mail: h796e@unb.ca; dlister@unb.ca; fsteward@unb.ca

    2005-07-01

    'Full text:' In CANDU-6 reactors, the pressurized high-temperature coolant flows through 380 fuel channels passing horizontally through the core. Each end of a fuel channel has a stainless steel annular end-fitting connected to a carbon steel feeder pipe. The outlet coolant, which is at 310{sup o}C with up to 0.30 steam voidage, turns through 90{sup o} as it passes from flow in the annular end-fitting to pipe flow in the feeder via a Grayloc connector. Since 1996, several CANDU stations have reported excessive corrosion of their outlet feeder pipes; especially between the first metre, which consisted of single or double bends. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow accelerated corrosion. In order to understand the hydrodynamics of the coolant in the outlet feeders by flow-visualization, a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream annular end-fitting were fabricated. The feeder consisted of a 54 mm inside diameter acrylic pipe with a 73{sup o} bend, connecting to an acrylic simulation of a Grayloc flanged fitting and annular end-fitting. The annular end-fitting consisted of an inner pipe, 110 mm outer diameter, and an outer pipe, 150 mm inner diameter, both 190.7 cm long in length. The tests were performed with water and air at atmospheric pressure and room temperature. The maximum water volumetric flow rate was 19 L/s and the volume fraction of air varied from 0.05 to 0.56. The phase distributions within the feeder pipe and along the length of the annulus were investigated with a digital video recorder. Size, concentration and velocity of the air bubbles at particular locations were studied with a high-speed digital still camera and a high-speed digital video camera. Phase distributions and variations in bubble size with velocity were determined. Particular attention was paid to the flow pattern at the inside

  12. Realistic thermal transient margin analysis of 'MONJU' based on plant performance measurements. Reactor vessel outlet nozzle and evaporator feed water inlet tube sheet of the manual reactor plant trip

    International Nuclear Information System (INIS)

    Yamada, Fumiaki; Mori, Takero

    2005-01-01

    In order to develop technologies and achieve safe and stable operation of Monju' as well as realize optimized design and construction of safe and economically competitive fast breeder reactors, the authors are evaluating design approach applied to 'Monju' based on actually measured behavioral data during plant operations. This report uses actual measured characteristic data of 'Monju' during a plant trip test obtained at a commissioning stage with up to 40% power output and introduces plant thermal hydraulic behavior analysis in a representative thermal transient event, i.e. a manual plant trip. Thermal transient driven loads incurred by the reactor vessel outlet nozzle and by the evaporator feed water inlet tube sheet were further derived by structural analyses and were compared with the previously derived values in the design stage and with the limit values. Though the reactor vessel outlet nozzle was exposed to larger temperature change in the trip test than the analytical prediction, the newly calculated mechanical load was about 50% of the previous evaluation in the design stage. Also, the newly analyzed mechanical load incurred by the evaporator feed water inlet tube sheet in this event had a large margin against the limit value of cumulative damage cycle fraction, although the observed temperature disturbance in a steam blow test was wilder than the analytical prediction. Thus we concluded that the Monju' plant has an assured safety margin against thermal transient in plant trip events. (author)

  13. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yoon, K. H.; Lee, C. B.

    2014-01-01

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness

  14. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  15. 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.)

  16. Experimental and computational studies of thermal mixing in next generation nuclear reactors

    Science.gov (United States)

    Landfried, Douglas Tyler

    The Very High Temperature Reactor (VHTR) is a proposed next generation nuclear power plant. The VHTR utilizes helium as a coolant in the primary loop of the reactor. Helium traveling through the reactor mixes below the reactor in a region known as the lower plenum. In this region there exists large temperature and velocity gradients due to non-uniform heat generation in the reactor core. Due to these large gradients, concern should be given to reducing thermal striping in the lower plenum. Thermal striping is the phenomena by which temperature fluctuations in the fluid and transferred to and attenuated by surrounding structures. Thermal striping is a known cause of long term material failure. To better understand and predict thermal striping in the lower plenum two separate bodies of work have been conducted. First, an experimental facility capable of predictably recreating some aspects of flow in the lower plenum is designed according to scaling analysis of the VHTR. Namely the facility reproduces jets issuing into a crossflow past a tube bundle. Secondly, extensive studies investigate the mixing of a non-isothermal parallel round triple-jet at two jet-to-jet spacings was conducted. Experimental results were validation with an open source computational fluid dynamics package, OpenFOAMRTM. Additional care is given to understanding the implementation of the realizable k-a and Launder Gibson RSM turbulence Models in OpenFOAMRTM. In order to measure velocity and temperature in the triple-jet experiment a detailed investigation of temperature compensated hotwire anemometry is carried out with special concern being given to quantify the error with the measurements. Finally qualitative comparisons of trends in the experimental results and the computational results is conducted. A new and unexpected physical behavior was observed in the center jet as it appeared to spread unexpectedly for close spacings (S/Djet = 1.41).

  17. 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.

  18. 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.

  19. CFD Study of Industrial FCC Risers: The Effect of Outlet Configurations on Hydrodynamics and Reactions

    Directory of Open Access Journals (Sweden)

    Gabriela C. Lopes

    2012-01-01

    Full Text Available Fluid catalytic cracking (FCC riser reactors have complex hydrodynamics, which depend not only on operating conditions, feedstock quality, and catalyst particles characteristics, but also on the geometric configurations of the reactor. This paper presents a numerical study of the influence of different riser outlet designs on the dynamic of the flow and reactor efficiency. A three-dimensional, three-phase flow model and a four-lump kinetic scheme were used to predict the performance of the reactor. The phenomenon of vaporization of the liquid oil droplets was also analyzed. Results showed that small changes in the outlet configuration had a significant effect on the flow patterns and consequently, on the reaction yields.

  20. Reactor container

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  1. Method of controlling the water quality in nuclear reactors

    International Nuclear Information System (INIS)

    Ibe, Hidefumi.

    1985-01-01

    Purpose: To obtain a simple and reliable water quality calculation system and water quality control method based thereon for the entire primary coolant circuits in BWR type reactors. Method: In a method of controlling the water quality of the reactor water by injecting hydrogen into the primary coolant circuits of a nuclear reactor, by utilizing a first linear relationship established between the concentration of oxygen and hydrogen in the main steam system and the concentration of radiolysis products in the reactor core and separators and mixing plenum portions, each of the above-mentioned concentrations is calculated from the concentrations for hydrogen or oxygen. Further, by utilizing the first linear relationship established between the concentrations for the oxygen and hydrogen in the recycling system and the concentration of the radiolysis products in the system from the downcomer to the lower plenum portion, the above-mentioned concentration is calculated from the concentration for oxygen and hydrogen. Then, the hydrogen injection rate into the primary coolant system is determined such that the calculated value takes an aimed value. (Ikeda, J.)

  2. Spiral-shaped disinfection reactors

    KAUST Repository

    Ghaffour, Noreddine

    2015-08-20

    This disclosure includes disinfection reactors and processes for the disinfection of water. Some disinfection reactors include a body that defines an inlet, an outlet, and a spiral flow path between the inlet and the outlet, in which the body is configured to receive water and a disinfectant at the inlet such that the water is exposed to the disinfectant as the water flows through the spiral flow path. Also disclosed are processes for disinfecting water in such disinfection reactors.

  3. Control of reactor coolant flow path during reactor decay heat removal

    International Nuclear Information System (INIS)

    Hunsbedt, A.N.

    1988-01-01

    This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

  4. Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward for 750–800°C Reactor Outlet Temperature

    Energy Technology Data Exchange (ETDEWEB)

    John Collins

    2009-08-01

    This document presents the NGNP Critical PASSCs and defines their technical maturation path through Technology Development Roadmaps (TDRMs) and their associated Technology Readiness Levels (TRLs). As the critical PASSCs advance through increasing levels of technical maturity, project risk is reduced and the likelihood of within-budget and on-schedule completion is enhanced. The current supplier-generated TRLs and TDRMs for a 750–800°C reactor outlet temperature (ROT) specific to each supplier are collected in Appendix A.

  5. Nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-05-10

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

  6. Nuclear reactors

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  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. 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)

  9. 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

  10. Control of ZrH reactor reactivity perturbations during orbital maneuvers

    International Nuclear Information System (INIS)

    Audette, R.F.

    1970-01-01

    Scheduled and inadvertent vehicle maneuvers in manned and unmanned space missions may result in reactivity perturbations to the ZrH reactor due to fuel and control drum motion from acceleration forces. Potential power and outlet coolant temperature excursions could result in interruptions of PCS power generation, or excessive coolant temperatures if uncontrolled. This analysis compares potential uncontrolled reactor transients with allowable transients for uninterrupted electrical power generation from a Brayton system, and presents a control scheme to limit transient reactor outlet temperatures to 1250 0 F for a system designed to operate at a nominal 1200 0 F reactor outlet. Potential uncontrolled transients could result in a reactor outlet temperature swing of +-77 0 F about a nominal 1200 0 F and a reactor power swing of +92 Kwt and -67 Kwt about a nominal 130 Kwt for the Brayton System. (U.S.)

  11. Bioconversion reactor

    Science.gov (United States)

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

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

  12. After-heat removal system of fast reactor

    International Nuclear Information System (INIS)

    Otsuka, Masaya; Shibata, Yoji; Ikeda, Takashi; Iwashige, Kengo; Yoneda, Yoshiyuki.

    1988-01-01

    Purpose: To remove after-heat by natural convection without disposing a movable portion even in a large-scaled reactor. Constitution: The exit of a reactor wall air-cooling duct disposed to the outside of a safety vessel is connected to the secondary inlet of an air cooler that conducts heat exchange with sodium in a high temperature plenum. That is, after-heat is removed only through the natural convection by a structure in which the reactor wall air-cooling duct and the secondary side of the air cooler are connected in series. Air exhausted from the exit of the air-cooling duct by the air cooler is further heated with sodium in the high temperature plenum. The flow rate of air flowing through the air-cooling duct is increased as compared with the case where the air cooler is not present. Accordingly, the flow rate of air at low temperature flowing through the inlet of the air duct is increased to increase the heat conduction amount. In this way, after-heat can be removed only by means of natural convection without providing movable portions even in a large-scaled reactor with the thermal power in excess of 2,000 MW. (Horiuchi, T.)

  13. Steam explosions in light water reactors

    International Nuclear Information System (INIS)

    1981-01-01

    The report deals with a postulated accident caused by molten fuel falling into the lower plenum of the containment of a reactor. The analysis which is presented in the report shows that the thermal energy released in the resulting steam explosion is not enough to destroy the pressure vessel or the containment. The report was prepared for the Swedish Governmental Committee on steam explosion in light water reactors. It includes statements issued by internationally well-known specialists. (G.B.)

  14. 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)

  15. 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

  16. Liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Scott, D.

    1981-01-01

    An improved method of constructing the diagrid used to support fuel assemblies of liquid metal fast breeder reactors, is described. The functions of fuel assembly support and coolant plenum are performed by discrete components of the diagrid each of which can serve the function of the other in the event of failure of one of the components. (U.K.)

  17. Savannah River Site reactor hardware design modification study

    International Nuclear Information System (INIS)

    Fisher, J.E.

    1990-03-01

    A study was undertaken to assess the merits of proposed design modifications to the SRS reactors. The evaluation was based on the responses calculated by the RELAP5 systems code to double-ended guillotine break loss-of-coolant-accidents (DEGB LOCAs). The three concepts evaluated were (a) elevated plenum inlet piping with a guard vessel and clamshell enclosures, (b) closure of both rotovalves in the affected loop, and (c) closure of the pump suction valve in the affected loop. Each concept included a fast reactor shutdown (to 65% power in 100 ms) and a 2-s ac pump trip. For the elevated piping design, system recovery was predicted for breaks in the plenum inlet or pump suction piping; response to the pump discharge break location did not show improvement compared to the present system configuration. The rotovalve closure design improved system response to plenum inlet or pump discharge breaks; recovery was not predicted for pump suction breaks. The pump suction valve closure design demonstrated system recovery for all break locations downstream of the valve. A combination of features is recommended to ensure liquid inventory recovery for all break locations. The elevated piping design performance during pump discharge breaks would be improved with addition of a dc pump trip in the affected loop. Valve closure design performance for a break location in the short section of piping between the reactor concrete shield and the pump suction valve would benefit from the clamshell enclosing that section of piping. 12 refs., 10 figs., 2 tabs

  18. 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.

  19. Computer supervision of the core outlet sodium temperatures of FBTR

    International Nuclear Information System (INIS)

    Boopathy, C.

    1976-01-01

    Safety monitoring of the fast breeder test reactor at Kalpakkam (India) is achieved by a CDPS-on-line dual computer system which is dedicated to plant supervision. The on-line subsystem scans and supervises all the 170 core thermocouple signals every second. Organisation of the reactor core instruments, supervision of mean sodium outlet temperature and mean temperature drop across the core, detection of plugging of a fuel assembly are explained. (A.K.)

  20. FBR type reactors

    International Nuclear Information System (INIS)

    Suzuoki, Akira; Yamakawa, Masanori.

    1985-01-01

    Purpose: To enable safety and reliable after-heat removal from a reactor core. Constitution: During ordinary operation of a FBR type reactor, sodium coolants heated to a high temperature in a reactor core are exhausted therefrom, collide against the reactor core upper mechanisms to radially change the flowing direction and then enter between each of the guide vanes. In the case if a main recycling pump is failed and stopped during reactor operation and the recycling force is eliminated, the swirling stream of sodium that has been resulted by the flow guide mechanism during normal reactor operation is continuously maintained within a plenum at a high temperature. Accordingly, the sodium recycling force in the coolant flow channels within the reactor vessel can surely be maintained for a long period of time due to the centrifugal force of the sodium swirling stream. In this way, since the reactor core recycling flow rate can be secured even after the stopping of the main recycling pump, after-heat from the reactor core can safely and surely be removed. (Seki, T.)

  1. Numerical simulation study on the air/water countercurrent flow limitation in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Morghi, Youssef; Mesquita, Amir Z., E-mail: ssfmorghi@gmail.com, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Puente, Jesus, E-mail: jpuente720@gmail.com [Centro Federal de Educaçao Tecnologica Celso Suckowda Fonseca (CEFET), Angra dos Reis, RJ (Brazil); Baliza, Ana R., E-mail: baliza@eletronuclear.gov.br [Eletrobras Eletronuclear Angra dos Reis, RJ (Brazil)

    2017-07-01

    After a loss-of-coolant accident (LOCA) in a Pressurized Water Reactor (PWR), the temperature of the fuel elements cladding increases dramatically due to the heat produced by the fission products decay, which is not adequately removed by the vapor contained in the core. In order to avoid this sharp rise in temperature and consequent melting of the core, the Emergency Core Cooling System is activated. This system initially injects borated water from accumulator tanks of the reactor through the inlet pipe (cold leg) and the outlet pipe (hot leg), or through the cold leg only, depending on the plant manufacturer. Some manufacturers add to this, direct injection into the upper plenum of the reactor. The penetration of water into the reactor core is a complex thermo fluid dynamic process because it involves the mixing of water with the vapor contained in the reactor, added to that generated in the contact of the water with the still hot surfaces in various geometries. In some critical locations, the vapor flowing in the opposite direction of the water can control the penetration of this into the core. This phenomenon is known as Countercurrent Flow Limitation (CCFL) or Flooding, and it is characterized by the control that a gas exerts in the liquid flow in the opposite direction. This work presents a proposal to use a CFD to simulate the CCFL phenomenon. Numerical computing can provide important information and data that is difficult or expensive to measure or test experimentally. Given the importance of computational science today, it can be considered a third and independent branch of science on an equal footing with the theoretical and experimental sciences. (author)

  2. 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

  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. Natural circulation in reactor coolant system

    International Nuclear Information System (INIS)

    Han, J.T.

    1987-01-01

    Reactor coolant system (RCS) natural circulation in a PWR is the buoyancy-driven coolant circulation between the core and the upper-plenum region (in-vessel circulation) with or without a countercurrent flow in the hot leg piping between the vessel and steam generators (ex-vessel circulation). This kind of multidimensional bouyancy-driven flow circulation serves as a means of transferring the heat from the core to the structures in the upper plenum, hot legs, and possibly steam generators. As a result, the RCS piping and other pressure boundaries may be heated to high temperatures at which the structural integrity is challenged. RCS natural circulation is likely to occur during the core uncovery period of the TMLB' accident in a PWR when the vessel upper plenum and hot leg are already drained and filled with steam and possibly other gaseous species. RCS natural circulation is being studied for the Surry plant during the TMLB' accident in which station blackout coincides with the loss of auxiliary feedwater and no operator actions. The effects of the multidimensional RCS natural circulation during the TMLB' accident are discussed

  6. 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

  7. 3D modeling of the primary circuit in the reactor pressure vessel of a PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramajo, Damian, E-mail: dramajo@santafe-conicet.gov.ar; Corzo, Santiago; Schiliuk, Nicolas; Nigro, Norberto

    2013-12-15

    A computational fluid dynamics (CFD) simulation of the reactor pressure vessel (RPV) of the pressurized heavy water reactor (PHWR) of 745 electrical MW Atucha II nuclear power plant was carried out. A three dimensional (3D) detailed model was employed to simulate coolant circuit considering the upper and lower plenums, the downcomer and the hot and cold legs. Control rods and coolant channel tubes at the upper plenum were included to quantify the mixing flow with more realism. The whole set of 451 coolant channels were modeled by means of a zero dimensional methodology. That is, the effect of each coolant channel was modeled through the introduction of a source point at the upper plenum and a sink point at the lower plenum. For each coupled sink/source points (SSP) the mass, momentum and energy balance were solved considering the local pressure difference and the temperature between the corresponding points where sinks and sources were placed. Based on this strategy, three models with increasingly level of approximation were implemented. For the first model the 451 coolant channels were reduced to only 57 pairs of SSP to represent all the coolant channels, concentrating the effect of several coolant channels in a unique pair of sink and source while taking into account geometric design details. For the second model, 225 pairs of SSP were introduced. Finally, for the third model each one of the 451 coolant channels were modeled by means of one pair of SSP. Depending on the coolant channel location, the radial power distribution and the pressure loss caused by the corresponding flow restrictor present by design were considered. Simulations carried out give insight in the complexity of the flow. As expected, the greater the details of the model the better the accuracy reached in the representation of the RPV behavior. In addition, the flow distributor located at the lower plenum showed to be very efficient since, the mass flow at each channel was found to be fairly

  8. Flow visualization study of two-phase flow in a single bend outlet feeder pipe of a CANDU reactor

    International Nuclear Information System (INIS)

    Savalaxs, S.-A.; Lister, D.H.; Steward, F.R.

    2005-01-01

    In CANDU reactors, the feeder piping that is used to direct the high-temperature water coolant between the fuel channels and the steam generators is made of carbon steel. Since 1996, several CANDU stations have reported excessive corrosion of their outlet feeders. The first metre is particularity vulnerable because the piping there consists of single or double bends, which have relatively thin walls produced by the bending process. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow-accelerated corrosion. In order to understand the hydrodynamics of the coolant in the outlet feeders by flow visualization, a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream components was fabricated. The feeder consisted of a 54 mm diameter acrylic pipe with a 73 degree bend. This was connected to the upstream component with an acrylic simulation of a Grayloc flanged fitting. A test loop supplied room temperature water to the test section at flow rates up to 0.019 m3/s. Air could be injected into the water to give a mean volume fraction of up to 0.56. In this preliminary investigation, the size and velocity of air bubbles at different flow conditions and their distribution within the pipe bend were studied. Particular attention was paid to the flow pattern at the inside of the bend, where a CFD (computational fluid dynamics) code - Fluent 6.1-had failed to predict a liquid film in an earlier study. A high-speed digital video camera was used to determine the relation between bubble size and velocity. Such a relation should help to explain the discrepancy in the CFD modelling and provide the basis for accurate predictions of phase distribution in complex geometries at high flow rates. (authors)

  9. Consideration of hot channel factors in design for providing operating margins on coolant channel outlet temperature

    International Nuclear Information System (INIS)

    Sharma, V.K.; Surendar, C.; Bapat, C.N.

    1994-01-01

    The Indian Pressurized Heavy Water Reactors (IPHWR) are horizontal pressure tube reactors using natural uranium oxide fuel in the form of short (495 mm) clusters. The fuel clusters in the Zr-Nb pressure tubes are cooled by high pressure, high temperature and subcooled circulating heavy water. Coolant flow distribution to individual channels is designed to match the power distribution so as to obtain uniform coolant outlet temperature. However, during operation, the coolant outlet temperature in individual channels deviate from their nominal value due to: tolerances in process design; effects of grid frequency on the pump speed; deviation in channel powers from the nominal values due to on-power fuelling and movement of reactivity devices, and so on. Thus an operating margin, between the highest permissible and nominal coolant outlet temperatures, is required taking into account various hot channel factors that contribute to higher coolant outlet temperatures. The paper discusses the methodology adopted to assess various hot channel factors which would provide optimum operating margins while ensuring sub-cooling. (author)

  10. 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

  11. 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.

  12. The DSNP simulation language and its application to liquid-metal fast breeder reactor transient analyses

    International Nuclear Information System (INIS)

    Saphier, D.; Madell, J.T.

    1982-01-01

    A new, special purpose block-oriented simulation language, the Dynamic Simulator for Nuclear Power Plants (DSNP), was used to perform a dynamic analysis of several conceptual design studies of liquid metal fast breeder reactors. The DSNP being a high level language enables the user to transform a power plant flow chart directly into a simulation program using a small number of DSNP statements. In addition to the language statements, the DSNP system has its own precompiler and an extensive library containing models of power plant components, algorithms of physical processes, material property functions, and various auxiliary functions. The comparative analysis covered oxide-fueled versus metal-fueled core designs and loop- versus pool-type reactors. The question of interest was the rate of change of the temperatures in the components in the upper plenum and the primary loop, in particular the reactor outlet nozzle and the intermediate heat exchanger inlet nozzle during different types of transients. From the simulations performed it can be concluded that metal-fueled cores will have much faster temperature transients than oxide-fueled cores due mainly to the much higher thermal diffusivity of the metal fuel. The transients in the pool-type design (either with oxide fuel or metal fuel) will be much slower than in the loop-type design due to the large heat capacity of the sodium pool. The DSNP language was demonstrated to be well suited to perform many types of transient analysis in nuclear power plants

  13. EBR-II [Experimental Breeder Reactor-II] system surveillance using pattern recognition software

    International Nuclear Information System (INIS)

    Mott, J.E.; Radtke, W.H.; King, R.W.

    1986-02-01

    The problem of most accurately determining the Experimental Breeder Reactor-II (EBR-II) reactor outlet temperature from currently available plant signals is investigated. Historically, the reactor outlet pipe was originally instrumented with 8 temperature sensors but, during 22 years of operation, all these instruments have failed except for one remaining thermocouple, and its output had recently become suspect. Using pattern recognition methods to compare values of 129 plant signals for similarities over a 7 month period spanning reconfiguration of the core and recalibration of many plant signals, it was determined that the remaining reactor outlet pipe thermocouple is still useful as an indicator of true mixed mean reactor outlet temperature. Application of this methodology to investigate one specific signal has automatically validated the vast majority of the 129 signals used for pattern recognition and also highlighted a few inconsistent signals for further investigation

  14. Numerical investigation of the coolant mixing during fast deboration transients for VVER-440 type reactors

    International Nuclear Information System (INIS)

    Hoehne, T.; Rhode, U.

    2000-01-01

    The VVER-440 (440 MW) V-230 was considered for analyzing the flow field and mixing processes. The V-230 has no elliptical sieve plate in the lower plenum. Previously, the 3D flow distribution in the downcomer and the lower plenum of the VVER-440 reactor have been calculated by means of the CFD code CFX-4 for operational conditions. The CFX-calculations were compared with the experimental data and the analytical mixing model. In this paper, CFD calculations for the start-up of the first main coolant pump in a VVER-440 type reactor are reported about. This scenario is important in case that there is a plug of lower borated water in one of the primary coolant loops. (orig.)

  15. Noise and DC balanced outlet temperature signals for monitoring coolant flow in LMFBR fuel elements

    International Nuclear Information System (INIS)

    Edelmann, M.

    1977-01-01

    Local cooling disturbances in LMFBR fuel elements may have serious safety implications for the whole reactor core. They have to be detected reliably in an early stage of their formation therefore. This can be accomplished in principle by individual monitoring of the coolant flow rate or the coolant outlet temperature of the sub-assemblies with high precision. In this paper a method is proposed to increase the sensitivity of outlet temperature signals to cooling disturbances. Using balanced temperature signals provides a means for eliminating the normal variations from the original signals which limit the sensitivity and speed of response to cooling disturbances. It is shown that a balanced signal can be derived easily from the original temperature signal by subtracting an inlet temperature and a neutron detector signal with appropriate time shift. The method was tested with tape-recorded noise signals of the KNK I reactor at Karlsruhe. The experimental results confirm the theoretical predictions. A significant reduction of the uncertainty of measured outlet temperatures was achieved. This enables very sensitive and fast response monitoring of coolant flow. Furthermore, it was found that minimizing the variance of the balanced signal offers the possibility for a rough determination of the heat transfer coefficient of the fuel rods during normal reactor operation at power. (author)

  16. Investigation on in-vessel thermal transients in a fast breeder reactor

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Kasahara, Naoto

    1999-01-01

    Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of the phenomena in the design of the internal structures in an LMFBR plenum. To evaluate thermal stress characteristics for the inner barrel in a typical LMFBR upper plenum, numerical analysis was carried out with a multi-dimensional thermohydraulics code AQUA for a scram condition from full power operation conditions. Thereafter, thermal stress conditions for the inner barrel were evaluated by the use of a structural analysis code FINAS with the thermohydraulic results calculated by the AQUA code as boundary conditions. From the thermohydraulic analysis and the thermal stress analysis, the following results have been obtained. (1) A large axial temperature gradient was calculated at the region between the upper and lower flow holes located on the inner barrel. The axial position of the thermal stratification interface was fixed in the various circumferential directions. As for the comparison with a 40% operation condition, maximum temperature gradients at the lower flow hole region indicated a 2 times value of that in the 40% operation condition. (2) Transient thermal stratification phenomena were observed after 120 sec from the reactor scram in the numerical results. These tendencies on thermal stratification phenomena were sameness with the transient results from the 40% operation condition. (3) During the reactor trip from full power operation, large temperature gradient in both vertical and sectional direction are enforced around the lower flow hole, since there exists flow pass of low temperature sodium through this hole. As a result, the maximum thermal stress within 32.6 kg/mm 2 was predicted at the lower flow hole when considering stress concentration at the hole edge. (J.P.N.)

  17. Requirements under decree 430 UJD for unit outlet and standby power supply

    International Nuclear Information System (INIS)

    Vanco, K.

    2012-01-01

    At present, the nuclear power plants are only sufficient resources, which can cover a huge demand for electricity. Concentration so huge power in one place require adequate security from the perspective lead power outlet and standby power supply of reactor unit. (Author)

  18. Method of injecting cooling water in emergency core cooling system (ECCS) of PWR type reactor

    International Nuclear Information System (INIS)

    Sobajima, Makoto; Adachi, Michihiro; Tasaka, Kanji; Suzuki, Mitsuhiro.

    1979-01-01

    Purpose: To provide a cooling water injection method in an ECCS, which can perform effective cooling of the reactor core. Method: In a method of injecting cooling water in an ECCS as a countermeasure against a rupture accident of a pwr type reactor, cooling water in the first pressure storage injection system is injected into the upper plenum of the reactor pressure vessel at a set pressure of from 50 to 90 atg. and a set temperature of from 80 to 200 0 C, cooling water in the second pressure storage injection system is injected into the lower plenum of the reactor pressure vessel at a pressure of from 25 to 60 atg. which is lower than the set pressure and a temperature less than 60 0 C, and further in combination with these procedures, cooling water of less than 60 0 C is injected into a high-temperature side piping, in the high-pressure injection system of upstroke of 100 atg. by means of a pump and the low-pressure injection system of upstroke of 20 atg. also by means of a pump, thereby cooling the reactor core. (Aizawa, K.)

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

    Science.gov (United States)

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

    2001-08-21

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

  20. Study on thermodynamic cycle of high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Qu Xinhe; Yang Xiaoyong; Wang Jie

    2017-01-01

    The development trend of the (very) High temperature gas-cooled reactor is to gradually increase the reactor outlet temperature. The different power conversion units are required at the different reactor outlet temperature. In this paper, for the helium turbine direct cycle and the combined cycle of the power conversion unit of the High temperature gas-cooled reactor, the mathematic models are established, and three cycle plans are designed. The helium turbine direct cycle is a Brayton cycle with recuperator, precooler and intercooler. In the combined cycle plan 1, the topping cycle is a simple Brayton cycle without recuperator, precooler and intercooler, and the bottoming cycle is based on the steam parameters (540deg, 6 MPa) recommended by Siemens. In the combined cycle plan 2, the topping cycle also is a simple Brayton cycle, and the bottoming cycle which is a Rankine cycle with reheating cycle is based on the steam parameters of conventional subcritical thermal power generation (540degC, 18 MPa). The optimization results showed that the cycle efficiency of the combined cycle plan 2 is the highest, the second is the helium turbine direct cycle, and the combined cycle plan 2 is the lowest. When the reactor outlet temperature is 900degC and the pressure ratio is 2.02, the cycle efficiency of the combined cycle plan 2 can reach 49.7%. The helium turbine direct cycle has a reactor inlet temperature above 500degC due to the regenerating cycle, so it requires a cooling circuit for the internal wall of the reactor pressure vessel. When the reactor outlet temperature increases, the increase of the pressure ratio required by the helium turbine direct cycle increases may bring some difficulties to the design and manufacture of the magnetic bearings. For the combined cycle, the reactor inlet temperature can be controlled below than 370degC, so the reactor pressure vessel can use SA533 steel without cooling the internal wall of the reactor pressure vessel. The pressure

  1. Auxiliary reactor for a hydrocarbon reforming system

    Science.gov (United States)

    Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

    2006-01-17

    An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

  2. 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.

  3. Nuclear reactor

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  4. 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

  5. Nuclear reactor cavity floor passive heat removal system

    Science.gov (United States)

    Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

    2018-03-06

    A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluid communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.

  6. Solvent refined coal reactor quench system

    Science.gov (United States)

    Thorogood, Robert M.

    1983-01-01

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream.

  7. Control Rod Driveline Reactivity Feedback Model for Liquid Metal Reactors

    International Nuclear Information System (INIS)

    Kwon, Young-Min; Jeong, Hae-Yong; Chang, Won-Pyo; Cho, Chung-Ho; Lee, Yong-Bum

    2008-01-01

    The thermal expansion of the control rod drivelines (CRDL) is one important passive mitigator under all unprotected accident conditions in the metal and oxide cores. When the CRDL are washed by hot sodium in the coolant outlet plenum, the CRDL thermally expands and causes the control rods to be inserted further down into the active core region, providing a negative reactivity feedback. Since the control rods are attached to the top of the vessel head and the core attaches to the bottom of the reactor vessel (RV), the expansion of the vessel wall as it heats will either lower the core or raise the control rods supports. This contrary thermal expansion of the reactor vessel wall pulls the control rods out of the core somewhat, providing a positive reactivity feedback. However this is not a safety factor early in a transient because its time constant is relatively large. The total elongated length is calculated by subtracting the vessel expansion from the CRDL expansion to determine the net control rod expansion into the core. The system-wide safety analysis code SSC-K includes the CRDL/RV reactivity feedback model in which control rod and vessel expansions are calculated using single-nod temperatures for the vessel and CRDL masses. The KALIMER design has the upper internal structures (UIS) in which the CRDLs are positioned outside the structure where they are exposed to the mixed sodium temperature exiting the core. A new method to determine the CRDL expansion is suggested. Two dimensional hot pool thermal hydraulic model (HP2D) originally developed for the analysis of the stratification phenomena in the hot pool is utilized for a detailed heat transfer between the CRDL mass and the hot pool coolant. However, the reactor vessel wall temperature is still calculated by a simple lumped model

  8. Development of 4S and related technologies (7). Analytical evaluation of RVACS performance under loss of stacks condition

    International Nuclear Information System (INIS)

    Nishimura, Satoshi; Nishi, Yoshihisa; Ueda, Nobuyuki

    2009-01-01

    The 4S (Super-Safe, Small and Simple) reactor is a sodium-cooled fast reactor aiming at an application to dispersed energy source and multi-purpose use. An introduction of RVACS (Reactor Vessel Auxiliary Cooling System) can enhance the passive decay heat removal capability. In the present study, the RVACS performance in 4S reactor (10MWe, pool-type) was analytically evaluated under the functional loss of RVACS stack condition, which is considered as a beyond design basis event. A protected loss of heat sink accident was selected and simulated to evaluate the capability of RVACS to cool the plant under such an unusual condition. The three-dimensional thermal hydraulic analysis was conducted by PHOENICS code. Analytical results show that the functional loss of air outlet stack has more effect on RVACS performance than that of air inlet stack. The air flow rate in RVACS under the functional loss of one out of two outlet stacks decreases up to approximately 60% and the heat removal rate approximately 70%, comparing with those under the normal stack condition. However, the maximum hot plenum temperature is low enough to satisfy the safety criteria. (author)

  9. Safe new reactor for radionuclide production

    International Nuclear Information System (INIS)

    Gray, P.L.

    1995-01-01

    In late 1995, DOE is schedule to announce a new tritium production unit. Near the end of the last NPR (New Production Reactors) program, work was directed towards eliminating risks in current designs and reducing effects of accidents. In the Heavy Water Reactor Program at Savannah River, the coolant was changed from heavy to light water. An alternative, passively safe concept uses a heavy-water-filled, zircaloy reactor calandria near the bottom of a swimming pool; the calandria is supported on a light-water-coolant inlet plenum and has upflow through assemblies in the calandria tubes. The reactor concept eliminates or reduces significantly most design basis and severe accidents that plague other deigns. The proven, current SRS tritium cycle remains intact; production within the US of medical isotopes such as Mo-99 would also be possible

  10. 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.

  11. Double Outlet Right Ventricle

    Science.gov (United States)

    ... Right Ventricle Menu Topics Topics FAQs Double Outlet Right Ventricle Double outlet right ventricle (DORV) is a rare form of congenital heart disease. En español Double outlet right ventricle (DORV) is a rare form of congenital ...

  12. 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)

  13. LWR type reactor

    International Nuclear Information System (INIS)

    Kato, Kiyoshi.

    1993-01-01

    A water injection tank in an emergency reactor core cooling system is disposed at a position above a reactor pressure vessel. A liquid phase portion of the water injection tank and an inlet plenum portion in the reactor pressure vessel are connected by a water injection pipe. A gas phase portion of the water injection tank and an upper portion in the reactor pressure vessel are connected by a gas ventilation pipe. Hydraulic operation valves are disposed in the midway of the water injection pipe and the gas ventilation pipe respectively. A pressure conduit is disposed for connecting a discharge port of a main recycling pump and the hydraulic operation valve. In a case where primary coolants are not sent to the main recycling pump by lowering of a liquid level due to loss of coolants or in a case where the main recycling pump is stopped by electric power stoppage or occurrence of troubles, the discharge pressure of the main recycling pump is lowered. Then, the hydraulic operation valve is opened to release the flow channel, then, boric acid water in the water injection tank is sent into the reactor by a falling head, to lead the reactor to a scram state. (I.N.)

  14. New finite element-based modeling of reactor core support plate failure

    Energy Technology Data Exchange (ETDEWEB)

    Pandazis, Peter; Lovasz, Liviusz [Gesellschaft fuer Anlagen- und Reaktorsicherheit gGmbH, Garching (Germany). Forschungszentrum; Babcsany, Boglarka [Budapest Univ. of Technology and Economics, Budapest (Hungary). Inst. of Nuclear Techniques; Hajas, Tamas

    2017-12-15

    ATHLET-CD is the severe accident module of the code system AC{sup 2} that is designed to simulate the core degradation phenomena including fission product release and transport in the reactor circuit, as well as the late phase processes in the lower plenum. In case of a severe accident degradation of the reactor core occurs, the fuel assemblies start to melt. The evolution of such processes is usually accompanied with the failure of the core support plate and relocation of the molten core to the lower plenum. Currently, the criterion for the failure of the support plate applied by ATHLET-CD is a user-defined signal which can be a specific time or process variable like mass, temperature, etc. A new method, based on FEM approach, was developed that could lead in the future to a more realistic criterion for the failure of the core support plate. This paper presents the basic idea and theory of this new method as well as preliminary verification calculations and an outlook on the planned future development.

  15. 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

  16. Flow visualization study of two-phase flow in the horizontal annulus of the fuel-channel outlet end-fitting of a CANDU reactor

    International Nuclear Information System (INIS)

    Supa-Amornkul, S.; Steward, F.R.; Lister, D.H.

    2005-01-01

    In CANDU-6 reactors, the pressurized hightemperature coolant flows through 380 fuel channels passing horizontally through the core. In 1996, higher than expected rates of wall thinning of the outlet feeders were ascribed to flow-accelerated corrosion (FAC). Such corrosion is strongly influenced by the hydrodynamics of the coolant. Results of preliminary flow visualization and modelling studies have suggested that flow conditions in the end-fitting annulus upstream of the outlet feeder may influence the pattern of FAC. For a full-scale flow visualization, an acrylic test section was built to simulate the cylindrical end-fitting with its annulus flow path. The tests were performed with water and air at atmospheric pressure and room temperature. The phase distribution along the length of the annulus was recorded with a digital video recorder. Size, concentration and velocity of the air bubbles at particular locations were studied with a high-speed digital still camera and a high-speed digital video camera. Phase distributions and variations in bubble size with velocity were determined. Significant effects on the flow patterns of spacer buttons in the annulus were observed. A commercial computational fluid dynamics (CFD) code-Fluent 6.1-was used to model the results. (authors)

  17. 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.

  18. 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.

  19. HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

    Science.gov (United States)

    Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

    1960-06-14

    A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

  20. 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.

  1. Development of special tools for the cleaning of reactor's interior in HANARO

    International Nuclear Information System (INIS)

    Cho, Y.-G.; Le, J.-H.; Ryu, J.-S.; Wu, J.-S.; Jung, H.-S.

    1999-01-01

    The HANARO (Hi-flux Advanced Neutron Application Reactor) in Korea has been being operated for 5 years, including one year of non-nuclear system commissioning tests since the installation of the reactor in early 1994. The HANARO is an open-tank-in-pool type reactor which has the advantage of free access from the pool top. The HANARO reactor had special cleaning works twice to remove debris from the inside reactor. This paper summarizes the development of special tools for reactor cleaning and how the reactor's inside had been successfully cleaned within short periods. The first cleaning work, after the initial flushing of the reactor system in early 1994, was the removal of the silica-gel sands, contaminated during installation, from the reactor pool and all equipment in the pool, including the reactor structure, the reactivity control units and the primary cooling system. Water-jet, pump suction, vacuum suction and whirl methods were used in combination with specially designed tools. The second one, occurred in February 1997 after two years of reactor operation was the cleaning work for the reactor's interior to remove several metal pieces broken from the parts of a check valve assembly in the primary cooling system. This work required development of many special tools that are all compact in size and remotely operable to reach all areas of the inlet plenum through very limited access holes. The special tools used for this work were two kinds of underwater cameras equipped with lighting, a debris-picking tool named 'revolving dustpan', two kinds of flow tube replacement tools and many other supplementary tools. All work had been successfully accomplished on the in-pool-platform temporarily installed 9m above the pool bottom to maintain the pool water level required in view of radiation shielding. Finally, the reactor internals were inspected using the underwater cameras to confirm the absence of debris and the surface integrity of the plenum as well as all fuel

  2. Optimization geometries of a vortex gliding-arc reactor for partial oxidation of methane

    International Nuclear Information System (INIS)

    Guofeng, Xu; Xinwei, Ding

    2012-01-01

    The effects of the geometry of gliding-arc reactor – such as distance between the electrodes, outlet diameter, and inlet position – on the reactor characteristics (methane conversion, hydrogen yield, and energy efficiency) have not been fully investigated. In this paper, AC gliding-arc reactors including the vortex flow configuration are designed to produce hydrogen from the methane by partial oxidation. The influence of vortex flow configuration on the reactor characteristics is also studied by varying the inlet position. When the inlet of the gliding-arc reactor is positioned close to the outlet, reverse vortex flow reactor (RVFR), the maximum energy efficiency reaches 50% and the yields of hydrogen and carbon monoxide are 40% and 65%, respectively. As the distance between electrodes increases from 5 mm to 15 mm, both hydrogen yield and energy efficiency increase approximately 10% for the RVFR. The energy efficiency and hydrogen yield are highest when the ratio of the outlet diameter to the inner diameter is 0.5 for the RVFR. Experimental results indicate that the flow field in the plasma reactor has an important influence on the reactor performance. Furthermore, hydrogen production increases as the number of feed gas flows in contact with the plasma zone increases. -- Highlights: ► Gliding-arc reactors were designed to produce hydrogen for studying the characteristics of the vortex flow reactor. ► Hydrogen yield of reverse vortex flow reactor was 10% higher than that of forward vortex flow reactor. ► Maximum energy efficiency was 50% for reverse vortex flow reactor. ► If discharge power was supplied to the reactors, the reactor performance increased with increasing distance between electrodes. ► Optimum ratio of the outlet and inner diameter was 1/2.

  3. 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

  4. 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

  5. Seals in nuclear reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The aim of this invention is the provision of improved seals for reactor vessels in which fuel assemblies are located together with inlets and outlets for the circulation of a coolant. The object is to provide a seal arrangement for the rotatable plugs of nuclear reactor closure heads which has good sealing capacities over a wide gap during operation of the reactor but which also permits uninhibited rotation of the plugs for maintenance. (U.K.)

  6. Reactor noise analysis of experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Ohtani, Hideji; Yamamoto, Hisashi

    1980-01-01

    As a part of dynamics tests in experimental fast reactor ''JOYO'', reactor noise tests were carried out. The reactor noise analysis techniques are effective for study of plant characteristics by determining fluctuations of process signals (neutron signal, reactor inlet temperature signals, etc.), which are able to be measured without disturbances for reactor operations. The aims of reactor noise tests were to confirm that no unstable phenomenon exists in ''JOYO'' and to gain initial data of the plant for reference of the future data. Data for the reactor noise tests treated in this paper were obtained at 50 MW power level. Fluctuations of process signals were amplified and recorded on analogue tapes. The analysis was performed using noise code (NOISA) of digital computer, with which statistical values of ASPD (auto power spectral density), CPSD (cross power spectral density), and CF (coherence function) were calculated. The primary points of the results are as follows. 1. RMS value of neutron signal at 50 MW power level is about 0.03 MW. This neutron fluctuation is not disturbing reactor operations. 2. The fluctuations of A loop reactor inlet temperatures (T sub(AI)) are larger than the fluctuations of B loop reactor inlet temperature (T sub(BI)). For this reason, the major driving force of neutron fluctuations seems to be the fluctuations of T sub(AI). 3. Core and blanket subassemblies can be divided into two halves (A and B region), with respect to the spacial motion of temperature in the reactor core. A or B region means the region in which sodium temperature fluctuations in subassembly are significantly affected by T sub(AI) or T sub(BI), respectively. This phenomenon seems to be due to the lack of mixing of A and B loop sodium in lower plenum of reactor vessel. (author)

  7. Supercritical Water Reactor Cycle for Medium Power Applications

    International Nuclear Information System (INIS)

    BD Middleton; J Buongiorno

    2007-01-01

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency (ge)20%; Steam turbine outlet quality (ge)90%; and Pumping power (le)2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

  8. Neutronic analysis of two-fluid thorium molten salt reactor

    International Nuclear Information System (INIS)

    Frybort, Jan; Vocka, Radim

    2009-01-01

    The aim of this paper is to evaluate features of the two-fluid MSBR through a parametric study and compare its properties to one-fluid MSBR concepts. The starting point of the analysis is the original ORNL 1000 MWe reactor design, although simplified to some extent. We studied the influence of dimensions of distinct reactor parts - fuel and fertile channels radius, plenum height, design etc. - on fundamental reactor properties: breeding ratio and doubling time, reactor inventory, graphite lifetime, and temperature feedback coefficients. The calculations were carried out using MCNP5 code. Based on obtained results we proposed an improved reactor design. Our results show clear advantages of the concept with two separate fluoride salts if compared to the one fluid concept in breading, doubling time, and temperature feedback coefficients. Limitations of the two-fluid concept - particularly the graphite lifetime - are also pointed out. The reactor design can be a subject of further optimizations, namely from the viewpoint of reactor safety. (author)

  9. Alcohol outlets, social disorganization, and robberies: accounting for neighborhood characteristics and alcohol outlet types.

    Science.gov (United States)

    Snowden, Aleksandra J; Freiburger, Tina L

    2015-05-01

    We estimated spatially lagged regression and spatial regime models to determine if the variation in total, on-premise, and off-premise alcohol outlet(1) density is related to robbery density, while controlling for direct and moderating effects of social disorganization.(2) Results suggest that the relationship between alcohol outlet density and robbery density is sensitive to the measurement of social disorganization levels. Total alcohol outlet density and off-premise alcohol outlet density were significantly associated with robbery density when social disorganization variables were included separately in the models. However, when social disorganization levels were captured as a four item index, only the association between off-premise alcohol outlets and robbery density remained significant. More work is warranted in identifying the role of off-premise alcohol outlets and their characteristics in robbery incidents. Copyright © 2015 Elsevier Inc. All rights reserved.

  10. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  12. SASSYS LMFBR systems analysis code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.

    1982-01-01

    The SASSYS code provides detailed steady-state and transient thermal-hydraulic analyses of the reactor core, inlet and outlet coolant plenums, primary and intermediate heat-removal systems, steam generators, and emergency shut-down heat removal systems in liquid-metal-cooled fast-breeder reactors (LMFBRs). The main purpose of the code is to analyze the consequences of failures in the shut-down heat-removal system and to determine whether this system can perform its mission adequately even with some of its components inoperable. The code is not plant-specific. It is intended for use with any LMFBR, using either a loop or a pool design, a once-through steam generator or an evaporator-superheater combination, and either a homogeneous core or a heterogeneous core with internal-blanket assemblies

  13. MIT pebble bed reactor project

    Energy Technology Data Exchange (ETDEWEB)

    Kadak, Andrew C. [Massachusetts Institute of Technology, Cambridge (United States)

    2007-03-15

    The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis.

  14. MIT pebble bed reactor project

    International Nuclear Information System (INIS)

    Kadak, Andrew C.

    2007-01-01

    The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis

  15. Device for preventing cooling water from flowing out of reactor

    International Nuclear Information System (INIS)

    Chinen, Masanori; Kotani, Koichi; Murase, Michio.

    1976-01-01

    Object: To provide emergency cooling system, which can prevent cooling water bearing radioactivity from flowing to the outside of the reactor at the time of breakage of feedwater pipe, thus eliminating the possibility of exposure of the fuel rod to provide high reliability and also reducing the possibility of causing radioactive pollution. Structure: The device for preventing cooling water from flowing out from the reactor features a jet nozzle inserted in a feedwater pipe adjacent to the inlet or outlet thereof immediately before the reactor container. The nozzle outlet is provided in the vicinity of the reactor wall and in a direction opposite to the direction of out-flow, and water supplied from a high pressure pump is jetted from it. (Nakamura, S.)

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

    International Nuclear Information System (INIS)

    Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi

    2002-08-01

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

  17. Recirculation system for nuclear reactors

    International Nuclear Information System (INIS)

    Braun, H. E.; Dollard, W. J.; Tower, S. N.

    1980-01-01

    A recirculation system for use in pressurized water nuclear reactors to increase the output temperature of the reactor coolant, thereby achieving a significant improvement in plant efficiency without exceeding current core design limits. A portion of the hot outlet coolant is recirculated to the inlets of the peripheral fuel assemblies which operate at relatively low power levels. The outlet temperature from these peripheral fuel assemblies is increased to a temperature above that of the average core outlet. The recirculation system uses external pumps and introduces the hot recirculation coolant to the free space between the core barrel and the core baffle, where it flows downward and inward to the inlets of the peripheral fuel assemblies. In the unlikely event of a loss of coolant accident, the recirculation system flow path through the free space and to the inlets of the fuel assemblies is utilized for the injection of emergency coolant to the lower vessel and core. During emergency coolant injection, the emergency coolant is prevented from bypassing the core through the recirculation system by check valves inserted into the recirculation system piping

  18. 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.)

  19. TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

    International Nuclear Information System (INIS)

    Gallardo, J.; Marquino, W.; Mistreanu, A.; Yang, J.

    2015-09-01

    The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

  20. TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, J. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Marquino, W.; Mistreanu, A.; Yang, J., E-mail: euqrop@hotmail.com [General Electric Hitachi Nuclear Energy, Wilmington, 28401 North Carolina (United States)

    2015-09-15

    The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

  1. Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    In, Wang Kee; Shin, Chang Hwan; Lee, Chan; Chun, Tae Hyun; Oh, Dong Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Chi Young [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2016-12-15

    The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermalhydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermalhydraulic technology and the commercialization.

  2. 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

  3. Simulation of mixing effects in a VVER-1000 reactor

    International Nuclear Information System (INIS)

    Ulrich Bieder; Gauthier Fauchet; Sylvie Betin; Nikola Kolev; Dimitar Popov

    2005-01-01

    Full text of publication follows: The work presented has been performed in the framework of the OECD/NEA thermalhydraulic benchmark V1000CT-2. This benchmark is related to fluid mixing in the reactor vessel during a MSLB accident scenario in a VVER-1000 reactor. The purpose of the first exercise is to test the capability of CFD codes to represent the coolant mixing in the reactor vessel, in particular in the downcomer and the lower plenum. Coolant mixing in a VVER-1000 V320 reactor was investigated in plant experiments during the commissioning of Kozloduy Unit 5 and 6. Starting from nearly symmetric states, asymmetric loop operation in different combinations was caused by disturbing the steam flow from one or more steam generators. Non-uniform and asymmetric loop flow mixing in the reactor vessel has been observed in the event of asymmetric loop operation. For certain flow patterns there is a shift (swirl) of the main loop flows with respect to the cold leg axes. This azimuthal shift as well as mixing coefficients from cold legs to the fuel assembly inlets have been measured. The presented reference problem is a pure TH problem with given boundary conditions and power distributions. During a stabilization phase, the thermal power of the reactor was 281 MW i.e. 9.36% of the nominal power according to primary balance. Then, a transient was initiated by closing the steam isolation valve of the steam generator one (SG-1) and isolating SG-1 from feed water. The coolant temperature in the cold and hot legs of Loop no 1 rose by 13-13.5 C. After about 20 minutes a stabilized state was reached which is considered as 'final state'. This final state has been analysed with the Trio-U code. Trio-U is a CFD code developed by the CEA Grenoble, aimed to supply an efficient computational tool to simulate transient thermalhydraulic mono-phase turbulent flows encountered in nuclear systems as well as in industrial processes. For the presented study, a LES approach was used. Therefore

  4. Pre evaluation for heat balance of prototype sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Han, Ji Woong; Kim, De Hee; Yoon, Jung; Kim, Eui Kwang; Lee, Tae Ho

    2012-01-01

    Under the long term advanced SFR R and D plan, the design of prototype reactor has been carried out toward the construction of the prototype SFR plant by 2028. The R and D efforts in fluid system design will be focused on developing a prototype design of primary heat transport system(PHTS), intermediate heat transport system (IHTS), decay heat removal system(DHRS), steam generation system(SGS), and related auxiliary system design for a prototype reactor as shown in Fig. 1. In order to make progress system design, top tier requirements for prototype reactor related to design parameters of NSSS and BOP should be decided at first. The top tier requirement includes general design basis, capacity and characteristics of reactor, various requirements related to safety, performance, securities, economics, site, and etc.. Extensive discussion has been done within Korea Atomic Energy Research Institute(KAERI) for the decision of top tier requirements of the prototype reactor. The core outlet temperature, which should be described as top tier requirements, is one of the critical parameter for system design. The higher core exit temperature could contribute to increase the plant efficiency. However, it could also contribute to decrease the design margin for structure and safety. Therefore various operating strategies based on different core outlet temperatures should be examined and evaluated. For the prototype reactor two core outlet temperatures are taken into accounted. The lower temperature is for the operation condition and the higher temperature is for the system design and licensing process of the prototype reactor. In order to evaluate the operability of prototype reactor designed based on higher temperature, the heat balance calculations have been performed at different core outlet temperature conditions. The electrical power of prototype reactor was assumed to be 100MWe and reference operating conditions were decided based on existing available data. The

  5. 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)

  6. 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)

  7. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1984-12-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  8. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1985-09-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  9. 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)

  10. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  11. Natural convection as the way of heat removal from fast reactor core at cooldown regimes

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Kuzina, J.A.; Uhov, V.A.; Sorokin, G.A.

    2000-01-01

    The problems of thermohydraulics in fast reactors at cooldown regimes at heat removal by natural convection are considered The results of experiments and calculations obtained in various countries in this area are presented. The special attention is given to heat removal through inter-assembly space in the core and also to problems of thermohydraulics in the upper plenum. (author)

  12. Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

    Directory of Open Access Journals (Sweden)

    Isaac Skavdahl

    2016-12-01

    Full Text Available Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX (Tco and the hot outlet temperature of the intermediate heat exchanger (Tho2 by manipulating the hot-side flow rates of the heat exchangers (Fh/Fh2 responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX (Tco only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1 flow rate manipulation; (2 reactor power manipulation; or (3 a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.

  13. Control of advanced reactor-coupled heat exchanger system: Incorporation of reactor dynamics in system response to load disturbances

    Energy Technology Data Exchange (ETDEWEB)

    Skavdahi, Isaac; Utgikar, Vivek [Dept. of Chemical and Materials Engineering, University of Idaho, Moscow (United States); Christensen, Richard [Nuclear Engineering Program, University of Idaho, Idaho Falls (United States); Chen, Ming Hui; Sun, Xiao Dong [Nuclear Engineering Program, Department of Mechanical and Aerospace Engineering, The Ohio State University, Columbus (United States); Sabharwall, Piyush [Idaho National Laboratory, Idaho Falls (United States)

    2016-12-15

    Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX (T{sub co}) and the hot outlet temperature of the intermediate heat exchanger (Th{sub o2}) by manipulating the hot-side flow rates of the heat exchangers (F{sub h}/F{sub h2}) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX (T{sub co}) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.

  14. Control of advanced reactor-coupled heat exchanger system: Incorporation of reactor dynamics in system response to load disturbances

    International Nuclear Information System (INIS)

    Skavdahi, Isaac; Utgikar, Vivek; Christensen, Richard; Chen, Ming Hui; Sun, Xiao Dong; Sabharwall, Piyush

    2016-01-01

    Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX (T_c_o) and the hot outlet temperature of the intermediate heat exchanger (Th_o_2) by manipulating the hot-side flow rates of the heat exchangers (F_h/F_h_2) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX (T_c_o) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change

  15. 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)

  16. Numerical study on coolant flow distribution at the core inlet for an integral pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Lin; Peng, Min Jun; Xia, Genglei; Lv, Xing; Li, Ren [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

    2017-02-15

    When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

  17. Pebble bed reactor with one-zone core

    International Nuclear Information System (INIS)

    Mueller-Frank, U.; Lohnert, G.

    1977-01-01

    The claim deals with measures to differentiate the flow rate and to remove spherical fuel elements in the core of a pebble bed reactor. Hence the vertical rate of the fuel elements in the border region is for example twice as much as in the centre. A central funnel-shaped outlet on the floor of the core container over which a conical body is placed with its peak pointing upwards, or also the forming of several outlets can be used to adjust to a certain exit rate for the fuel elements. The main target of the invention is a radially extensively constant coolant outlet temperature at the outlet of the core which determines the effectiveness of the connected heat exchanger and thus contributes to economy. (UA) [de

  18. Temperature measurements at the LMFBR core outlet

    International Nuclear Information System (INIS)

    Argous, J.P.; Berger, R.; Casejuane, R.; Fournier, C.; Girard, J.P.

    1980-04-01

    Over the last few years the temperature sensors used to measure the subassembly outlet temperature in French designed LMFBRs have been modified, basically in an effort to reduce the dispersion of the chromel-alumel thermocouple time constant, and to extend the frequency spectrum of the measurement signals by adding a steel electrode to from a stainless steel-sodium thermocouple. The result of this evolution is the temperature probe immersed in sodium which will be used in the SUPER PHENIX reactor. This paper describes the tests already completed or in progress on this probe. It also presents measurement data on the two basic probe parameters: the thermoelectric power of the stainless steel-sodium thermocouple and the time constant of the chromel-alumel thermocouple

  19. 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

  20. Reactor core design optimization of the 200 MWt Pb-Bi cooled fast reactor for hydrogen production

    International Nuclear Information System (INIS)

    Bahrum, Epung Saepul; Su'ud, Zaki; Waris, Abdul; Fitriyani, Dian; Wahjoedi, Bambang Ari

    2008-01-01

    In this study reactor core geometrical optimization of 200 MWt Pb-Bi cooled long life fast reactor for hydrogen production has been conducted. The reactor life time is 20 years and the fuel type is UN-PuN. Geometrical core configurations considered in this study are balance, pancake and tall cylindrical cores. For the hydrogen production unit we adopt steam membrane reforming hydrogen gas production. The optimum operating temperature for the catalytic reaction is 540degC. Fast reactor design optimization calculation was run by using FI-ITB-CHI software package. The design criteria were restricted by the multiplication factor that should be less than 1.002, the average outlet coolant temperature 550degC and the maximum coolant outlet temperature less than 700degC. By taking into account of the hydrogen production as well as corrosion resulting from Pb-Bi, the balance cylindrical geometrical core design with diameter and height of the active core of 157 cm each, the inlet coolant temperature of 350degC and the coolant flow rate of 7000 kg/s were preferred as the best design parameters. (author)

  1. Method of repairing pressure tube type reactors

    International Nuclear Information System (INIS)

    Asada, Takashi.

    1983-01-01

    Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

  2. 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

  3. Thermal-hydraulic modeling of porous bed reactors

    International Nuclear Information System (INIS)

    Araj, K.J.; Nourbakhsh, H.P.

    1987-01-01

    Optimum design of nuclear reactor core requires an iterative approach between the thermal-hydraulic, neutronic and operational analysis. This paper concentrates on the thermal-hydraulic behavior of a hydrogen cooled, small particle bed reactor (PBR). The PBR core, modeled here, consists of a hexagonal array of fuel elements embedded in a moderator matrix. The fuel elements are annular packed beds of fuel particles held between two porous cylindrical frits. These particles, 500 to 600 μm in diameter, have a uranium carbide core, which is coated by two layers of graphite and an outer coating of zirconium carbide. Coolant flow, radially inward, from the cold frit through the packed bed and hot frit and axially out the channel, formed by the hot frit, to a common plenum. 5 refs., 1 fig., 2 tabs

  4. Evaluation of the RELAP5/MOD3 multidimensional component model

    International Nuclear Information System (INIS)

    Tomlinson, E.T.; Rens, T.E.; Coffield, R.D.

    1994-01-01

    Accurate plenum predictions, which are directly related to the mixing models used, are an important plant modeling consideration because of the consequential impact on basic transient performance calculations for the integrated system. The effect of plenum is a time shift between inlet and outlet temperature changes to the particular volume. Perfect mixing, where the total volume interacts instantaneously with the total inlet flow, does not occur because of effects such as inlet/outlet nozzle jetting, flow stratification, nested vortices within the volume and the general three-dimensional velocity distribution of the flow field. The time lag which exists between the inlet and outlet flows impacts the predicted rate of temperature change experienced by various plant system components and this impacts local component analyses which are affected by the rate of temperature change. This study includes a comparison of two-dimensional plenum mixing predictions using CFD-FLOW3D, RELAP5/MOD3 and perfect mixing models. Three different geometries (flat, square and tall) are assessed for scalar transport times using a wide range of inlet velocity and isothermal conditions. In addition, the three geometries were evaluated for low flow conditions with the inlet flow experiencing a large step temperature decrease. A major conclusion from this study is that the RELAP5/MOD3 multidimensional component model appears to be adequately predicting plenum mixing for a wide range of thermal-hydraulic conditions representative of plant transients

  5. Design Guideline for Primary Heat Exchanger in a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sunil; Seo, Kyoung-Woo; Kim, Seong-Hoon; Chi, Dae-Young; Park, Cheol [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this paper, analytical study is conducted to track the variation of the PCS outlet temperature in conditions of the constant core power and constant SCS inlet temperature. The PCS circulates demineralized water to remove the heat generated in reactor core. The heat is transferred to the cold water of the SCS through the primary heat exchanger. In JRTR, Plate-type Heat Exchanger (PHE) was used as the primary heat exchanger. The cooling tower automatically sets the SCS inlet temperature constant by fan speed control. The flow rate of SCS is adjusted to be identical with the PCS flow rate. To design the PHE, the inlet and outlet temperatures and the flow rates for both systems should be determined. The flow rate has the allowable band for the safe operation from the lower limit to upper limit resulting in different temperature distribution in the PHE. Specially, the PCS outlet temperature which is the core inlet temperature is used for a safety parameter for the reactor shutdown. Therefore, we need to figure out which limit for the flow rate should be used from the conservative point of view. At 200 kg/s of PCS and SCS flow rates, the inlet and outlet temperatures are 41.3℃and 34℃, respectively. With increase of the flow rate, both of PCS inlet and outlet temperatures decrease to 33.6℃ and 39.9℃. This result means the low limit of the allowable flow band should be used for the conservative design of primary heat exchanger. If the upper limit of the allowable flow band is used, the PCS outlet temperature which is the safety parameter used for the reactor shutdown increases with decrease of the flow rate.

  6. Design Guideline for Primary Heat Exchanger in a Research Reactor

    International Nuclear Information System (INIS)

    Lee, Sunil; Seo, Kyoung-Woo; Kim, Seong-Hoon; Chi, Dae-Young; Park, Cheol

    2016-01-01

    In this paper, analytical study is conducted to track the variation of the PCS outlet temperature in conditions of the constant core power and constant SCS inlet temperature. The PCS circulates demineralized water to remove the heat generated in reactor core. The heat is transferred to the cold water of the SCS through the primary heat exchanger. In JRTR, Plate-type Heat Exchanger (PHE) was used as the primary heat exchanger. The cooling tower automatically sets the SCS inlet temperature constant by fan speed control. The flow rate of SCS is adjusted to be identical with the PCS flow rate. To design the PHE, the inlet and outlet temperatures and the flow rates for both systems should be determined. The flow rate has the allowable band for the safe operation from the lower limit to upper limit resulting in different temperature distribution in the PHE. Specially, the PCS outlet temperature which is the core inlet temperature is used for a safety parameter for the reactor shutdown. Therefore, we need to figure out which limit for the flow rate should be used from the conservative point of view. At 200 kg/s of PCS and SCS flow rates, the inlet and outlet temperatures are 41.3℃and 34℃, respectively. With increase of the flow rate, both of PCS inlet and outlet temperatures decrease to 33.6℃ and 39.9℃. This result means the low limit of the allowable flow band should be used for the conservative design of primary heat exchanger. If the upper limit of the allowable flow band is used, the PCS outlet temperature which is the safety parameter used for the reactor shutdown increases with decrease of the flow rate

  7. Quarterly technical progress report on water reactor safety programs sponsored by the Nuclear Regulatory Commission's Division of Reactor Safety Research, April--June 1977

    Energy Technology Data Exchange (ETDEWEB)

    1977-08-01

    Results from the previously conducted Semiscale Mod-1 ECC injection test series were analyzed. Testing in the LOFT counterpart test series was essentially completed, and the steam generator tube rupture test series was begun. Two tests in the alternate ECC injection test series were conducted which included injection of emergency core coolant into the upper plenum through use of the low pressure injection system. The Loss-of-Fluid Test Program successfully completed nonnuclear Loss-of-Coolant Experiment L1-4. A nuclear test, GC 2-3, in the Power Burst Facility Reactor was performed to evaluate the power oscillation method of determining gap conductance and to determine the effects of initial gap size, fill gas composition, and fuel density on the thermal performance of a light water reactor fuel rod. Additional test results were obtained relative to the behavior of irradiated fuel rods during a fast power increase and during a high power film boiling transient. Fuel model development and verification activities continued for the steady state and transient Fuel Rod Analysis Program, FRAP-S and FRAP-T. A computer code known as RELAP4/MOD7 is being developed to provide best-estimate modeling for reflood during a postulated loss-of-coolant accident (LOCA). A prediction of the fourth test in the boiling water reactor (BWR) Blowdown/Emergency Core Cooling Program was completed and an uncertainty analysis was completed of experimental steady state stable film boiling data for water flowing vertically upward in round tubes. A new multinational cooperative program to study the behavior of entrained liquid in the upper plenum and cross flow in the core during the reflood phase of a pressurized water reactor LOCA was defined.

  8. Experimental study on the safety of Kyoto University Research Reactor at natural circulation cooling mode

    International Nuclear Information System (INIS)

    Zhang, Jian; Shen, Xiuzhong; Fujihara, Yasuyuki; Sano, Tadafumi; Yamamoto, Toshihiro; Nakajima, Ken

    2015-01-01

    Highlights: • The natural circulation cooling capacity of Kyoto University Research Reactor (KUR) was experimentally investigated. • The distributions of the outlet temperature of the fuel elements under natural circulation operations were measured. • The average temperature rise and the average natural circulation flow velocity in core were calculated. • The safety of KUR under all of the normal operations with natural circulation cooling mode has been analyzed. • The natural circulation flow after the reactor shutdown was confirmed. - Abstract: In this study, the natural circulation cooling capacity of Kyoto University Research Reactor (KUR) is experimentally investigated by measuring the inlet and outlet temperatures of the core under natural circulation operation at various thermal powers ranging from 10 kW to 100 kW and the shutdown state. In view of the uneven power distribution and the resultant inconsistent coolant outlet temperature in the core, eight measuring points located separately in the outlet of the fuel elements were chosen to investigate the distribution of the outlet temperature of the core. The natural circulation cooling capacity represented by the average natural circulation flow velocity in the core is calculated from the temperature difference between the outlet and inlet temperature of the core. The measured outlet temperature of the fuel elements shows a cross-sectional distribution agreeing with the distribution of the thermal output of the fuel elements in the core. Since the measured outlet temperatures decrease quickly in the flow direction in a small local region above the outlet of the core, the mixing of the hot water out of the core with the cold water around the core outlet is found to happen in the small region not more than 5 cm far from the core outlet. The natural circulation flow velocity in the core increases non-linearly with the thermal power. The safety of KUR has been analysed by conservatively estimating the

  9. 14 CFR 23.977 - Fuel tank outlet.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank outlet. 23.977 Section 23.977... tank outlet. (a) There must be a fuel strainer for the fuel tank outlet or for the booster pump. This... damage any fuel system component. (b) The clear area of each fuel tank outlet strainer must be at least...

  10. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  11. Thermal-hydraulic mixing in the split-core ANS reactor design

    International Nuclear Information System (INIS)

    Dorning, R.J.J.

    1988-01-01

    A design has been proposed for the advanced neutron source (ANS) reactor that incorporates a split core, one purpose of which is to create a mixing plenum between the upper and lower cores. It was hoped that in addition to introducing various desirable neutronics features, such as decreasing the fast neutron flux contamination of thermal and cold neutron beams located in the reactor midplane, this mixing plenum would make possible higher operating powers by lowering the maximum core temperature. This lower temperature was to be achieved as a result of the mixing, of the hot D 2 O coolant exiting the upper-core channels, and the cold D 2 O leaving the large upper core bypass. It was expected that this mixing would bring about a significantly reduced lower core maximum coolant inlet temperature. The authors have carried out large-scale computer calculations to determine the extent to which this mixing occurs in current split-core design geometry, which does not incorporate baffles, mixing devices, or other design features introduced to enhance mixing. The large-scale self-consistent calculations summarized here indicate that innovative design ideas to enhance mixing will be necessary if the split-core concept is to achieve the amount of thermal mixing needed to make possible significantly higher power operation and corresponding higher flux sources

  12. Reactor safety issues resolved by the 2D/3D Program

    International Nuclear Information System (INIS)

    Damerell, P.S.; Simons, J.W.

    1993-07-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated

  13. Continuous Assessment of Safety Margin for the 14-MW TRIGA Reactor

    International Nuclear Information System (INIS)

    Ciocanescu, M.; Georgescu, D.; Doru, O.

    2008-01-01

    The assessment of reactor safety implies analyses of the reactor and its systems response to a range of postulated initiating events (such as malfunction or failures of equipment, operator errors, external events and so on which could lead to either anticipated operational occurrences or to accident conditions. Decreasing in heat removal by the reactor cooling system may be considered as a process disturbance which may lead to a postulated initiating event. The cold source for the reactor cooling system, in case of TRIGA-14 MW reactor is the atmosphere by the secondary cooling towers. The ability to evacuate the heat produced by the reactor core ranges between the outlet temperature of the core flow and the outdoors temperature in air, which is subject to season and day variation. Selected values for safety limits, safety system settings and limiting condition(s) are derived from safety analysis and are consistent with the operational state of the reactor. When a limiting condition for safe operation is not satisfied, the operating personal is supposed to take the appropriate action(s) to ensure safety. Operating requirements and the safety system are presented. The reactor operating safety parameters from the main Data Acquisition System are transferred to an AT personal computer. These selected parameters are the following: - average inlet temperature which is calculated as an average temperature measured by 20 type K thermocouples distributed within a 4 x 5 matrix located on the top of the reactor core; - average outlet temperature which is calculated as an average record from 10 type K thermocouples placed in the outlet pipe; - average flow rate which is calculated as an average value from four transducers (two for the inlet flow rate and two for the outlet flow rate). Due to its high instability, this value is also filtered using a two-pole low-pass filter (software); - reactor thermal power value derivable from the previous parameters or obtained from the

  14. Core design and fuel rod analyses of a super fast reactor with high power density

    International Nuclear Information System (INIS)

    Ju, Haitao; Cao, Liangzhi; Lu, Haoliang; Oka, Yoshiaki; Ikejiri, Satoshi; Ishiwatari, Yuki

    2009-01-01

    A Super Fast Reactor is a pressure-vessel type, fast spectrum SuperCritical Water Reactor (SCWR) that is presently researched in a Japanese project. One of the most important advantages of the Super Fast Reactor is the higher power density compared to the thermal spectrum SCWR, which reduces the capital cost. A preliminary core has an average power density of 158.8W/cc. In this paper, the principle of improving the average power density is studied and the core design is improved. After the sensitivity analyses on the fuel rod configurations, the fuel assembly configurations and the core configurations, an improved core with an average power density of 294.8W/cc is designed by 3-D neutronic/thermal-hydraulic coupled calculations. This power density is competitive with that of typical Liquid Metal Fast Breeder Reactors (LMFBR). In order to ensure the fuel rod integrity of this core design, the fuel rod behaviors on the normal operating condition are analyzed using FEMAXI-6 code. The power histories of each fuel rod are taken from the neutronics calculation results in the core design. The cladding surface temperature histories are taken from the thermal-hydraulic calculation results in the core design. Four types of the limiting fuel rods, with the Maximum Cladding Surface Temperature (MCST), Maximum Power Peak(MPP), Maximum Discharge Burnup(MDB) and Different Coolant Flow Pattern (DCFP), are chosen to cover all the fuel rods in the core. The available design range of the fuel rod design parameters, such as initial gas plenum pressure, gas plenum position, gas plenum length, grain size and gap size, are found out in order to satisfy the following design criteria: (1) Maximum fuel centerline temperature should be less than 1900degC. (2) Maximum cladding stress in circumstance direction should be less than 100MPa. (3) Pressure difference on the cladding should be less than 1/3 of buckling collapse pressure. (4) Cumulative damage faction (CDF) of the cladding should be

  15. Temperature etalon of WWER-440 reactor

    International Nuclear Information System (INIS)

    Stanc, S.; Slanina, M.

    2001-01-01

    The presentation deals with the description, parameters and advantages of use of the temperature etalon. The system ensures temperature measurement of reactor outlet and inlet temperatures with high accuracy. Accuracy of temperature measurement is 0.18 deg C, accuracy of temperature difference measurement is 0.14 deg C, both with probability 0.95. Using the temperature etalon it is possible to increase accuracy of the standard temperature reactor measurements and to check their accuracy in the course of power reactor statuses in every measurement cycle. Temperature reactor etalon was installed in 12 WWER-440 units in Slovakia, Bohemia and Bulgaria. (Authors)

  16. A study on different thermodynamic cycle schemes coupled with a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Qu, Xinhe; Yang, Xiaoyong; Wang, Jie

    2017-01-01

    Highlights: • The features of three different power generation schemes, including closed Brayton cycle, non-reheating combined cycle and reheating combined cycle, coupled with high temperature gas-cooled reactor (HTGR) were investigated and compared. • The effects and mechanism of reactor core outlet temperature, compression ratio and other key parameters over cycle characteristics were analyzed by the thermodynamic models.. • It is found that reheated combined cycle has the highest efficiency. Reactor outlet temperature and main steam parameters are key factors to improve the cycle’s performance. - Abstract: With gradual increase in reactor outlet temperature, the efficient power conversion technology has become one of developing trends of (very) high temperature gas-cooled reactors (HTGRs). In this paper, different cycle power generation schemes for HTGRs were systematically studied. Physical and mathematical models were established for these three cycle schemes: closed Brayton cycle, simple combined cycle, and reheated combined cycle. The effects and mechanism of key parameters such as reactor core outlet temperature, reactor core inlet temperature and compression ratio on the features of these cycles were analyzed. Then, optimization results were given with engineering restrictive conditions, including pinch point temperature differences. Results revealed that within the temperature range of HTGRs (700–900 °C), the reheated combined cycle had the highest efficiency, while the simple combined cycle had the lowest efficiency (900 °C). The efficiencies of the closed Brayton cycle, simple combined cycle and reheated combined cycle are 49.5%, 46.6% and 50.1%, respectively. These results provide insights on the different schemes of these cycles, and reveal the effects of key parameters on performance of these cycles. It could be helpful to understand and develop a combined cycle coupled with a high temperature reactor in the future.

  17. Pebble-bed reactor

    International Nuclear Information System (INIS)

    Lohnert, G.; Mueller-Frank, U.; Heil, J.

    1976-01-01

    A pebble-bed nuclear reactor of large power rating comprises a container having a funnel-shaped bottom forming a pebble run-out having a centrally positioned outlet. A bed of downwardly-flowing substantially spherical nuclear fuel pebbles is positioned in the container and forms a reactive nuclear core maintained by feeding unused pebbles to the bed's top surface while used or burned-out pebbles run out and discharge through the outlet. A substantially conical body with its apex pointing upwardly and its periphery spaced from the periphery of the container spreads the bottom of the bed outwardly to provide an annular flow down the funnel-shaped bottom forming the runout, to the discharge outlet. This provides a largely constant downward velocity of the spheres throughout the diameter of the bed throughout a substantial portion of the down travel, so that all spheres reach about the same burned-out condition when they leave the core, after a single pass through the core area

  18. Alcohol Outlet Density and Intimate Partner Violence in a Nonmetropolitan College Town: Accounting for Neighborhood Characteristics and Alcohol Outlet Types.

    Science.gov (United States)

    Snowden, Aleksandra J

    2016-01-01

    There is a growing evidence of an ecological association between alcohol outlet density and intimate partner violence. It is reasonable to assume, however, that not all types of alcohol outlets contribute equally to criminal behavior, and to date, most ecological studies have been of large urban cities. Using Bloomington, Indiana, block groups as units of analysis and controlling for several structural characteristics associated with violence rates, I estimated spatially lagged regression models to determine if the variation in alcohol outlet density, including total outlets and disaggregating by on- and off-premise outlets, is related to intimate partner violence density. Results suggested that total alcohol outlet density and off-premise alcohol outlet density were significantly associated with intimate partner violence density. On-premise alcohol outlet density was not significantly associated with intimate partner violence density. These results not only extend the geographic scope of this relationship beyond large metropolitan areas but also have important policy implications.

  19. Simplified simulation of an experimental fast reactor plant

    International Nuclear Information System (INIS)

    Fujii, Masaaki; Fujita, Minoru.

    1978-01-01

    Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)

  20. A thermal analysis computer programme package for the estimation of KANUPP coolant channel flows and outlet header temperature distribution

    International Nuclear Information System (INIS)

    Siddiqui, M.S.

    1992-06-01

    COFTAN is a computer code for actual estimation of flows and temperatures in the coolant channels of a pressure tube heavy water reactor. The code is being used for Candu type reactor with coolant flowing 208 channels. The simulation model first performs the detailed calculation of flux and power distribution based on two groups diffusion theory treatment on a three dimensional mesh and then channel powers, resulting from the summation of eleven bundle powers in each of the 208 channels, are employed to make actual estimation of coolant flows using channel powers and channel outlet temperature monitored by digital computers. The code by using the design flows in individual channels and applying a correction factor based on control room monitored flows in eight selected channels, can also provide a reserve computational tool of estimating individual channel outlet temperatures, thus providing an alternate arrangements for checking Rads performance. 42 figs. (Orig./A.B.)

  1. Improved reactor cavity

    International Nuclear Information System (INIS)

    Katz, L.R.; Demarchais, W.E.

    1984-01-01

    A reactor pressure vessel disposed in a cavity has coolant inlet or outlet pipes extending through passages in the cavity walls and welded to pressure nozzles. The cavity wall has means for directing fluid away from a break at a weld away from the pressure vessel, and means for inhibiting flow of fluid toward the vessel. (author)

  2. Validation and verification of the MTRPC thermohydraulic package

    International Nuclear Information System (INIS)

    Doval, Alicia

    1998-01-01

    The MTR P C v2.6 is a computational package developed for research reactor design and calculation. It covers three of the main aspects of a research reactor: neutronic, shielding and thermohydraulic. In this work only the thermohydraulic package will be covered, dealing with verification and validation aspects. The package consists of the following steady state programs: CAUDVAP 2.60 for the hydraulic calculus, estimates the velocity distribution through different parallel channels connected to a common inlet and outlet common plenum. TERMIC 1H v3.0, used for the thermal design of research reactors, provides information about heat flux for a given maximum wall temperature, onset of nucleate boiling, redistribution phenomena and departure from nucleate boiling. CONVEC V3.0 allows natural convection calculations, giving information on heat fluxes for onset of nucleate boiling, pulsed and burn-out phenomena as well as total coolant flow. Results have been validated against experimental values and verified against theoretical and computational programmes results, showing a good agreement. (author)

  3. Thermal hydraulic analysis of gas-cooled reactors with annular fuel rods

    International Nuclear Information System (INIS)

    Han, Kyu Hyun; Chang, Soon Heung

    2005-01-01

    More than half of the world's energy is used in industrial processes and for heating applications which have hardly been touched by the nuclear industry. Nuclear power could be brought into a wide range of applications for industrial processes, provided that gas outlet temperatures of gascooled reactors are sufficiently high. The most limiting core design requirement which controls the core outlet temperature is the maximum acceptable fuel compact temperature. An innovative fuel design is required for a significant decrease in the fuel temperature. This study investigated the possibilities of implementing internally and externally cooled annular fuel rods in a gas-cooled reactor

  4. Detailed flow analysis for the Three Mile Island unit 2 reactor accident

    International Nuclear Information System (INIS)

    Lillington, J.N.; Lyons, A.J.

    1990-01-01

    Some particular characteristics of the steam flow in the accident at the Three Mile Island unit 2 pressurized water reactor are investigated using the AEA Technology Flow3D code. Natural circulation flows with heat removal from the core and deposition in the upper plenum are predicted during the primary heating phase. The structure of the upper plenum cylinder and core blockage, owing to material relocation, are shown to force the flow into a complex three-dimensional pattern. The flows and temperature distributions from the calculations are shown to be consistent with the observed damage pattern above the core. Despite high core temperatures, damage was limited by the operation of one of the pumps at the end of the initial heating phase. Flow3D calculations are also carried out to demonstrate that the three-dimensional buoyancy driven flows are completely destroyed by the high steam generation rates arising from the pump operation. (author)

  5. 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

  6. 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

  7. Investigation of flow stabilization in a compact reactor vessel of a FBR. Flow visualization in a reactor vessel

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Igarashi, Minoru; Kimura, Nobuyuki; Kamide, Hideki

    2002-01-01

    In the feasibility studies of Commercialized Fast Breeder Reactor Cycle System, a compact reactor vessel is considered from economical improvement point of a sodium cooled loop type fast reactor. The flow field was visualized by water experiment for a reactor vessel with 'a column type UIS (Upper Internal Structure)', which has a slit for fuel handling mechanism and is useful for a compact fast reactor. In this research, the 1/20 scale test equipment using water was made to understand coolant flow through a slit of a column type UIS' and fundamental behavior of reactor upper plenum flow. In the flow visualization tests, tracer particles were added in the water, and illuminated by the slit-shaped pulse laser. The flow visualization image was taken with a CCD camera. We obtained fluid velocity vectors from the visualization image using the Particle Imaging Velocimetry (PIV). The results are as follows. 1. Most of coolant flow through a slit of 'column type UIS' arrived the dip plate directly. In the opposite side of a slit, most of coolant flowed toward reactor vessel wall before it arrived the dip plate. 2. The PIV was useful to measure the flow field in the reactor vessel. The obtained velocity field was consistent with the flow visualization result. 3. The jet through the UIS slit was dependent on the UIS geometry. There is a possibility to control the jet by the UIS geometry. (author)

  8. Support structure for reactor core constituent element

    International Nuclear Information System (INIS)

    Aida, Yasuhiko.

    1993-01-01

    A connection pipe having an entrance nozzle inserted therein as a reactor core constituent element is protruded above the upper surface of a reactor core support plate. A through hole is disposed to the protruding portion of the connection pipe. The through hole and a through hole disposed to the reactor core support plate are connected by a communication pipe. A shear rod is disposed in a horizontal portion at the inside of the communication pipe and is supported by a spring horizontally movably. Further, a groove is disposed at a position of the entrance nozzle opposing to the shear rod. The shear rod is urged out of the communication pipe by the pressure of the high pressure plenum and the top end portion of the shear rod is inserted to the groove of the entrance nozzle during operation. Accordingly, the shear rod is positioned in a state where it is extended from the through hole of the communication pipe to the groove of the entrance nozzle. This can mechanically constrain the rising of the reactor core constituent elements by the shear rod upon occurrence of earthquakes. (I.N.)

  9. Analysis of a Natural Circulation in the Reactor Coolant System Following a High Pressure Severe Accident at APR1400

    International Nuclear Information System (INIS)

    Kim, Han Chul; Cho, Yong Jin; Park, Jae Hong; Cho, Song Won

    2011-01-01

    Under a high temperature and pressure condition during a severe accident, hot leg pipes or steam generator tubes could fail due to creep rupture following natural circulation in the Reactor Coolant System (RCS) unless depressurization of the system is performed at a proper time. Natural circulation in the RCS can be a multi-dimensional circulation in the reactor vessel, a partial loop circulation of two-phase flow from the core up to steam generators (SGs), or circulation in the total loop. It can delay the reactor vessel failure time by removing heat from the reactor core. This natural phenomenon can be hardly simulated with a single flow path model for the hot spots of the RCS, since it cannot deal with the counter-current flow. Thus it may estimate accident progression faster than reality, which may cause troubles for optimized implementation of severe accident management strategies. An earlier damage in the RCS other than the reactor pressure vessel may make subsequent behaviors of hydrogen or fission products in the containment quite different from the single reactor vessel failure. Therefore, a RCS model which treats natural circulation is needed to evaluate the RCS response and the safety depressurization strategy in a best-estimate way. The aim of this study is to develop a detailed model which allows natural circulation between the reactor vessel and steam generators through hot legs, based on the existing APR1400 RCS model. The station blackout sequence was selected to be the representative high-pressure scenario. Sensitivity study on the effect of node configuration of the upper plenum and addition of cross flow paths from the upper plenum to the hot legs were carried out. This model is described herein and representative calculation results are presented

  10. Pressure loss coefficient evaluation based on CFD analysis for simple geometries and PWR reactor vessel without geometry simplification

    International Nuclear Information System (INIS)

    Ko II, B.; Park, J. P.; Jeong, J. H.

    2008-01-01

    Nuclear vendors and utilities perform lots of simulations and analyses in order to ensure the safe operation of nuclear power plants (NPPs). In general, the simulations are carried out using vendor-specific design codes and best-estimate system analysis codes and most of them were developed based on 1-dimensional lumped parameter models. These thermal-hydraulic system analysis codes require user input for pressure loss coefficient, k-factor; since they numerically solve Euler-equation. In spite of its high impact on the safety analysis results, there has not been good validation method for the selection of loss coefficient. During the past decade, however; computers, parallel computation methods, and 3-dimensional computational fluid dynamics (CFD) codes have been dramatically enhanced. It is believed to be beneficial to take advantage of advanced commercial CFD codes in safety analysis and design of NPP5. The present work aims to validate pressure loss coefficient evaluation for simple geometries and k-factor calculation for PWR based on CFD. The performances of standard k-ε model, RNG k-ε model, Reynolds stress model (RSM) on the simulation of pressure drop for simple geometry such as, or sudden-expansion, and sudden-contraction are evaluated. The calculated value was compared with pressure loss coefficient in handbook of hydraulic resistance. Then the present work carried out analysis for flow distribution in downcomer and lower plenum of Korean standard nuclear power plants (KSNPs) using STAR-CD. The lower plenum geometry of a PWR is very complicated since there are so many reactor internals, which hinders in CFD analysis for real reactor geometry up to now. The present work takes advantage of 3D CAD model so that real geometry of lower plenum is used. The results give a clear figure about flow fields in the reactor vessel, which is one of major safety concerns. The calculated pressure drop across downcomer and lower plenum appears to be in good agreement

  11. Socioeconomic determinants of exposure to alcohol outlets.

    Science.gov (United States)

    Morrison, Christopher; Gruenewald, Paul J; Ponicki, William R

    2015-05-01

    Alcohol outlets tend to be located in lower income areas, exposing lower income populations to excess risks associated with alcohol sales through these establishments. The objective of this study was to test two hypotheses about the etiology of these differential exposures based on theories of the economic geography of retail markets: (a) outlets will locate within or near areas of high alcohol demand, and (b) outlets will be excluded from areas with high land and structure rents. Data from the 2010 National Drug Strategy Household Survey were used to develop a surrogate for alcohol demand (i.e., market potential) at two census geographies for the city of Melbourne, Australia. Bayesian conditional autoregressive Poisson models estimated multilevel spatial relationships between counts of bars, restaurants, and off-premise outlets and market potential, income, and zoning ordinances (Level 1: n = 8,914). Market potentials were greatest in areas with larger older age, male, English-speaking, high-income populations. Independent of zoning characteristics, greater numbers of outlets appeared in areas with greater market potentials and the immediately surrounding areas. Greater income excluded outlets in local and surrounding areas. These findings are consistent with the hypothesis that alcohol outlets are located in areas with high demand and are excluded from high-income areas. These processes appear to take place at relatively small geographic scales, encourage the concentration of outlets in specific low-income areas, and represent a very general economic process likely to take place in communities throughout the world.

  12. Off-premise alcohol outlet characteristics and violence.

    Science.gov (United States)

    Snowden, Aleksandra J; Pridemore, William Alex

    2014-07-01

    There is considerable evidence of an association between alcohol outlet density and violence. Although prior research reveals the importance of specific characteristics of bars on this association and that the relationship between bar density and violence may be moderated by these characteristics, there are few similar studies of the characteristics of off-premise outlets (e.g., liquor and convenience stores). We examined whether immediate environment, business practice, staff, and patron characteristics of off-premise alcohol outlets are associated with simple and aggravated assault density. Cross-sectional design using aggregate data from 65 census block groups in a non-metropolitan college town, systematic social observation, and spatial modeling techniques. We found limited effects of immediate environment, business practice, staff, and patron characteristics on simple assault density and no effect on aggravated assault density. Only two out of 17 characteristics were associated with simple assault density (i.e., nearby library and male patrons). This is the first study to examine the association between several off-premise alcohol outlet characteristics and assault. Our findings suggest that where the off-premise outlets are located, how well the immediate environment is maintained, what types of beverages the outlets sell, who visits them, and who works there matter little in their association with violence. This suggests the importance of outlet density itself as a primary driver of any association with violence. Public policies aimed at reducing alcohol outlet density or clustering may be useful for reducing violence.

  13. 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.

  14. MELCOR development for existing and advanced reactors

    International Nuclear Information System (INIS)

    Summers, R.M.

    1993-01-01

    Recent efforts in MELCOR development to address previously identified deficiencies have resulted in release of MELCOR 1.8.2, a much-improved version of the code. Major new models have been implemented for direct containment heating, ice condensers, debris quenching, lower plenum debris behavior, core materials interactions' and radial relocation of debris. Significant improvements have also been made in the modeling of interfacial momentum exchange and in the modeling of fission product release, condensation/evaporation, and aerosol behavior. Efforts are underway to address two-phase hydrodynamics difficulties, to improve modeling of water condensation on structures and fine-scale natural circulation within the reactor vessel, and to implement CORCON-Mod3. Improvements are also being made to MELCOR's capability to handle new features of the advanced light water reactor designs, including drainage of water films on connected heat structures, heat transfer from the external surface of the reactor vessel to a flooded cavity, and creep rupture failure of the lower head. Additional development needs in other areas are discussed

  15. Steam conversion of liquefied petroleum gas and methane in microchannel reactor

    Science.gov (United States)

    Dimov, S. V.; Gasenko, O. A.; Fokin, M. I.; Kuznetsov, V. V.

    2018-03-01

    This study presents experimental results of steam conversion of liquefied petroleum gas and methane in annular catalytic reactor - heat exchanger. The steam reforming was done on the Rh/Al2O3 nanocatalyst with the heat applied through the microchannel gap from the outer wall. Concentrations of the products of chemical reactions in the outlet gas mixture are measured at different temperatures of reactor. The range of channel wall temperatures at which the ratio of hydrogen and carbon oxide in the outlet mixture grows substantially is determined. Data on the composition of liquefied petroleum gas conversion products for the ratio S/C = 5 was received for different GHVS.

  16. Design and evaluation of materials for space reactors

    International Nuclear Information System (INIS)

    Tavassoli, A.A.; Vrillon, B.; Robert, G.

    1990-01-01

    The French programme envisages a 20 kWe reactor, project ERATO, with three technological options. The first option is a sodium cooled reactor, derived from the fast breeder reactor technology, (upper core outlet temperature of 700 0 C). The second option is based on the High Temperature Gas-cooled Reactor technology (outlet temperature range 700 0 C-900 0 C). The third option is the reference solution, lithium cooled and UN fuelled fast spectrum reactor, (outlet temperature as high as 1200 0 C). The choice is essentially dominated by material considerations, and more specifically by the problems related to the compatibility with the cooling medium and to the high temperature creep resistance. For the first system limited work will be needed as the technology used is well experimented and there is a wealth of information on the austenitic stainless steel Type 316L-SPH. For the second system, most of the work has been concentrated on characterization of existing commercial alloys. This has included the preselection and the testing of a number of superalloys irradiated or not. The results obtained from high temperature tensile and creep tests have allowed selection of Haynes 230 as the primary candidate material and have also permitted calculation of allowable design stresses for this alloy. For the very high temperature system the French R and D programme has focused on Mo-Re alloys. The results obtained to this date from microstructural examinations and mechanical tests performed on different alloy compositions have allowed selection of Mo-25%Re for future optimization work. They have also shown the need for evaluation of creep properties at low stresses where microstructural instabilities are likely to occur as a result of long exposure to high temperature

  17. Sensitivity Studies of Advanced Reactors Coupled to High Temperature Electrolysis (HTE) Hydrogen Production Processes

    International Nuclear Information System (INIS)

    Edwin A. Harvego; Michael G. McKellar; James E. O'Brien; J. Stephen Herring

    2007-01-01

    High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the steam or air sweep loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycle producing the highest efficiencies varied depending on the temperature range considered

  18. 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

  19. Structural analysis of the Upper Internals Structure for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Houtman, J.L.

    1979-01-01

    The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides control of core outlet flow to prevent severe thermal transients from occuring at the reactor vessel and primary heat transport outlet piping, provides instrumentation to monitor core performance, provides support for the control rod drivelines, and provides secondary holddown of the core. All of the structural analysis aspects of assuring the UIS is structurally adequate are presented including simplified and rigorous inelastic analysis methods, elevated temperature criteria, environmental effects on material properties, design techniques, and manufacturing constraints

  20. 14 CFR 25.977 - Fuel tank outlet.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank outlet. 25.977 Section 25.977... STANDARDS: TRANSPORT CATEGORY AIRPLANES Powerplant Fuel System § 25.977 Fuel tank outlet. (a) There must be a fuel strainer for the fuel tank outlet or for the booster pump. This strainer must— (1) For...

  1. 14 CFR 29.977 - Fuel tank outlet.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank outlet. 29.977 Section 29.977... STANDARDS: TRANSPORT CATEGORY ROTORCRAFT Powerplant Fuel System § 29.977 Fuel tank outlet. (a) There must be a fuel strainer for the fuel tank outlet or for the booster pump. This strainer must— (1) For...

  2. 14 CFR 27.977 - Fuel tank outlet.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Fuel tank outlet. 27.977 Section 27.977... STANDARDS: NORMAL CATEGORY ROTORCRAFT Powerplant Fuel System § 27.977 Fuel tank outlet. (a) There must be a fuel stainer for the fuel tank outlet or for the booster pump. This strainer must— (1) For...

  3. Ultrasonic sweep arm for sodium cooled reactors

    International Nuclear Information System (INIS)

    Rohrbacher, H.A.; Bartholomay, R.

    1975-05-01

    This report describes experience in the use of a new type of monitoring and testing device to be applied in conjunction with components under sodium. In the method outlined, ultrasonic pulses are used which are emitted into the sodium plenum of fast breeder reactors by newly developed high temperature transducers. The basic work was conducted under out-of-pile conditions in a sodium tank of the sodium tank facility of the Karlsruhe Institute for Reactor Development. The sensor development, which preceded this phase, resulted in the use of soldered lithium niobate crystals whose operating characteristics were improved by the preliminary treatment outlined in the report. Special materials and techniques suitable for sensor fabrication are proposed. An alternative to soldering is suggested for contacting the crystals with their diaphragms, i.e. a contact pressure concept for the range of application up to 2 MHz. (orig.) [de

  4. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  5. Socioeconomic Determinants of Exposure to Alcohol Outlets

    Science.gov (United States)

    Morrison, Christopher; Gruenewald, Paul J.; Ponicki, William R.

    2015-01-01

    Objective: Alcohol outlets tend to be located in lower income areas, exposing lower income populations to excess risks associated with alcohol sales through these establishments. The objective of this study was to test two hypotheses about the etiology of these differential exposures based on theories of the economic geography of retail markets: (a) outlets will locate within or near areas of high alcohol demand, and (b) outlets will be excluded from areas with high land and structure rents. Method: Data from the 2010 National Drug Strategy Household Survey were used to develop a surrogate for alcohol demand (i.e., market potential) at two census geographies for the city of Melbourne, Australia. Bayesian conditional autoregressive Poisson models estimated multilevel spatial relationships between counts of bars, restaurants, and off-premise outlets and market potential, income, and zoning ordinances (Level 1: n = 8,914). Results: Market potentials were greatest in areas with larger older age, male, English-speaking, high-income populations. Independent of zoning characteristics, greater numbers of outlets appeared in areas with greater market potentials and the immediately surrounding areas. Greater income excluded outlets in local and surrounding areas. Conclusions: These findings are consistent with the hypothesis that alcohol outlets are located in areas with high demand and are excluded from high-income areas. These processes appear to take place at relatively small geographic scales, encourage the concentration of outlets in specific low-income areas, and represent a very general economic process likely to take place in communities throughout the world. PMID:25978830

  6. Safety analysis of an expert reactor protection system in nuclear power plants

    International Nuclear Information System (INIS)

    EL-Kafas, A.E.A.E.

    1996-01-01

    the purpose of the dissertation is to develop a real time expert reactor protection system (ERPS) for operational safety of pressurized water reactor nuclear power plant. The system is developed to diagnose plant failures and for identification of plant transients (with and without scram). for this ERPS. probabilistic safety analysis techniques are used to check the availability and priority of the recommended safety system in case of plant accidents . the real- time information during transients and accidents can be obtained to asses the operator in his decision - making . Also, the ERPS is able to give advice for the reactor operator to take the appropriate corrective action during abnormal situations. The system model consists of the dynamic differential equations for reactor core, pressurizer, steam generator, turbine and generator, piping and plenums. The system of equations can be solved by appropriate codes also displayed directly from sensors of the plant. All scenarios of transients, accidents and fault tress for plant systems are learned to ERPS

  7. Assessment of some interfacial shear correlations in a model of ECC bypass flow in PWR reactor downcomer

    International Nuclear Information System (INIS)

    Popov, N.K.; Rohatgi, U.S.

    1987-01-01

    The bypass/refill process in the PWR reactor downcomer, following a large rupture of a cold leg coolant supply pipe, is a complicated thermo-hydraulic two-phase flow phenomenon. Mathematical modeling of such phenomena is always accompanied with a difficult task of selection of suitable constitutive correlations. In a typically hydrodynamic phenomenon, like ECC refill process of the reactor lower plenum is considered, the phasic interfacial friction is the most influential constitutive correlation. Therefore, assessment of the well-known widely-used interfacial friction constitutive correlations in the model of ECC bypass/refill process, is the subject of this paper

  8. Unique features of space reactors

    International Nuclear Information System (INIS)

    Buden, D.

    1990-01-01

    This paper reports on space reactors that are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K

  9. Particle Bed Reactor scaling relationships

    International Nuclear Information System (INIS)

    Slovik, G.; Araj, K.; Horn, F.L.; Ludewig, H.; Benenati, R.

    1987-01-01

    Scaling relationships for Particle Bed Reactors (PBRs) are discussed. The particular applications are short duration systems, i.e., for propulsion or burst power. Particle Bed Reactors can use a wide selection of different moderators and reflectors and be designed for such a wide range of power and bed power densities. Additional design considerations include the effect of varying the number of fuel elements, outlet Mach number in hot gas channel, etc. All of these variables and options result in a wide range of reactor weights and performance. Extremely light weight reactors (approximately 1 kg/MW) are possible with the appropriate choice of moderator/reflector and power density. Such systems are very attractive for propulsion systems where parasitic weight has to be minimized

  10. Development of Alloy 718 tubular product for nuclear-reactor internals

    International Nuclear Information System (INIS)

    1981-01-01

    The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides mixing and flow direction of the core outlet flow. Alloy 718 tubes are the major components used in the UIS to provide this flow direction. The UIS is located directly above the reactor core and is exposed to a severe environment. This environment consists of high temperature sodium, alternating temperatures induced by mixing high temperature core assembly outlet flow with cooler core assembly outlet flow and rapid changes in temperature of the core outlet flow. The paper presents the UIS configuration, functions and environmental conditions that led to the selection of Alloy 718 as the material used to protect the basic UIS structure and to provide flow direction. The use of Alloy 718 is derived from the technology produced from the Department of Energy sponsored development programs. Alloy 718 tubes are made by a roll-extrusion process. The paper describes the tube fabrication process, the development of a finish sanding procedure and the results of high temperature thermal cycle testing. The high temperature thermal cycle testing generates peak strains on the metal surfaces, where the surface effects have a maximum influence on the fatigue strength. 9 figs., 2 tabs

  11. Reactor safety device

    International Nuclear Information System (INIS)

    Okada, Yasumasa.

    1987-01-01

    Purpose: To scram control rods by processing signals from a plurality of temperature detectors and generating abnormal temperature warning upon occurrence of abnormal temperature in a nuclear reactor. Constitution: A temperature sensor comprising a plurality of reactors each having a magnetic body as the magnetic core having a curie point different from each other and corresponding to the abnormal temperature against which reactor core fuels have to be protected is disposed in an identical instrumentation well near the reactor core fuel outlet/inlet of a reactor. A temperature detection device actuated upon detection of an abnormal temperature by the abrupt reduction of the reactance of each of the reactors is disposed. An OR circuit and an AND circuit for conducting OR and AND operations for each of the abnormal temperature detection signals from the temperature detection device are disposed. The output from the OR circuit is used as the abnormal temperature warning signal, while the output from the AND circuit is utilized as a signal for actuating the scram operation of control rod drive mechanisms. Accordingly, it is possible to improve the reliability of the reactor scram system, particularly, improve the reliability under a high temperature atmosphere. (Kamimura, M.)

  12. Complot test section outlet CFD optimization (pre - test and dimensioning)

    International Nuclear Information System (INIS)

    Profir, M. M.; Moreau, V.; Kennedy, G.

    2013-01-01

    In the framework of the FP7 MAXSIMA European project, the COMPLOT (COMPonent LOop Testing) LBE experimental facility is employed for thermal-hydraulic experiments aimed to test and qualify, among other components, a buoyancy driven safety/control rods (SR/CR) system, as key components for the safe operation of the MYRRHA reactor. This paper focuses mainly on a simplified CFD representation of the SR test section outlet in order to optimise it for the testing program. Parametric cases, associated with different positions of the SR assembly have been set up and analysed. A quasi-static analysis has been performed for each case, accounting for the LBE volume displaced by the insertion of the SR bundle, by introducing appropriately positioned additional mass sources. Velocity and pressure fields, as well as pressure drop magnitudes and mass flow rates through relevant guide tube hole outlets have been calculated and compared. The CFD analysis proved that the outer boundary of the test section does not impact the expected performance of the SR (rapid transient downward insertion). Preliminary simulations reproducing the timely repositioning of the SR/CR in COMPLOT using procedures of automatic volume mesh regeneration, consistently with the rod imposed displacement, are illustrated. (authors)

  13. Nuclear reactor with several cores

    International Nuclear Information System (INIS)

    Swars, H.

    1977-01-01

    Several sodium-cooled cores in separate vessels with removable closures are placed in a common reactor tank. Each individual vessel is protected against the consequences of an accident in the relevant core. Maintenance devices and inlet and outlet pipes for the coolant are also arranged within the reactor tank. The individual vessels are all enclosed by coolant in a way that in case of emergency cooling or refuelling each core can be continued to be cooled by means of the coolant loops of the other cores. (HP) [de

  14. Double-outlet right ventricle revisited.

    Science.gov (United States)

    Ebadi, Ameneh; Spicer, Diane E; Backer, Carl L; Fricker, F Jay; Anderson, Robert H

    2017-08-01

    Double-outlet right ventricle is a form of ventriculoarterial connection. The definition formulated by the International Society for Nomenclature of Paediatric and Congenital Heart Disease is based on hearts with both arterial trunks supported in their greater part by a morphologically right ventricle. Bilateral infundibula and ventricular septal defects are highly debated criteria. This study examines the anatomic controversies surrounding double-outlet right ventricle. We show that hearts with double-outlet right ventricle can have atrioventricular-to-arterial valvular continuity. We emphasize the difference between the interventricular communication and the zone of deficient ventricular septation. The hearts examined were from the University of Florida in Gainesville; Johns Hopkins All Children's Hospital, St Petersburg, Fla; and Lurie Children's Hospital, Chicago, Ill. Each specimen had at least 75% of both arterial roots supported by the morphologically right ventricle, with a total of 100 hearts examined. The morphologic method was used to assess anatomic features, including arterial-atrioventricular valvular continuity, subarterial infundibular musculature, and the location of the hole between the ventricles. Most hearts had fibrous continuity between one of the arterial valves and an atrioventricular valve, with bilateral infundibula in 23%, and intact ventricular septum in 5%. Bilateral infundibula are not a defining feature of double-outlet right ventricle, representing only 23% of the specimens in our sample. The interventricular communication can have a posteroinferior muscular rim or extend to become perimembranous (58%). Double-outlet right ventricle can exist with an intact ventricular septum. Copyright © 2017 The American Association for Thoracic Surgery. All rights reserved.

  15. Thermal-hydraulic analysis for the LBE-cooled natural circulation reactor. Development of the MSG-COPD code and application to the system analysis. Research Document

    International Nuclear Information System (INIS)

    Iwasaki, Takashi; Sakai, Takaaki; Enuma, Yasuhiro; Mizuno, Tomoyasu

    2002-03-01

    Thermal-hydraulic analysis for the Lead-Bismuth eutectic (LBE)-cooled natural circulation reactor has been conducted by using a combined plant dynamics code (MSG-COPD). MSG-COPD has been developed to consider the multi-dimensional thermal-hydraulics effect on the plant dynamics during transients. Plant dynamics analyses for the LBE-cooled STAR-LM reactor, which has been designed by Argonne National Laboratory in U.S.A., have been performed to understand the basic thermal-hydraulic characteristics of the natural circulation reactor. As a result, it has been made clear that cold coolant remains in the lower plenum by the thermal stratification in case of the ULOHS condition with a severe temperature gradient at the stratified surface in the lower plenum. In addition, the flow-redistribution effect in a core channels by the buoyancy force has been evaluated for a candidate LBE-cooled FBR plant concept (LBE-FR), which has been designed by JNC. A linear evaluation method for the flow-redistribution coefficient is proposed for the LBE-FR, and compared with the multi-dimensional results by MSG-COPD. In conclusion, the method shows sufficient performance for the prediction of the flow-redistribution coefficient for typical lateral power distributions in the core. (author)

  16. 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.

  17. Research program of the high temperature engineering test reactor for upgrading the HTGR technology

    International Nuclear Information System (INIS)

    Kunitomi, Kazuhiko; Tachibana, Yukio; Takeda, Takeshi; Saikusa, Akio; Sawa, Kazuhiro

    1997-07-01

    The High Temperature Engineering Test Reactor (HTTR) is a graphite-moderated and helium-cooled reactor with an outlet power of 30 MW and outlet coolant temperature of 950degC, and its first criticality will be attained at the end of 1997. In the HTTR, researches establishing and upgrading the technology basis necessary for an HTGR and innovative basic researches for a high temperature engineering will be conducted. A research program of the HTTR for upgrading the technology basis for the HTGR was determined considering realization of future generation commercial HTGRs. This paper describes a research program of the HTTR. (author)

  18. Intelligent electrical outlet for collective load control

    Science.gov (United States)

    Lentine, Anthony L.; Ford, Justin R.; Spires, Shannon V.; Goldsmith, Steven Y.

    2015-10-27

    Various technologies described herein pertain to an electrical outlet that autonomously manages loads in a microgrid. The electrical outlet can provide autonomous load control in response to variations in electrical power generation supply in the microgrid. The electrical outlet includes a receptacle, a sensor operably coupled to the receptacle, and an actuator configured to selectively actuate the receptacle. The sensor measures electrical parameters at the receptacle. Further, a processor autonomously controls the actuator based at least in part on the electrical parameters measured at the receptacle, electrical parameters from one or more disparate electrical outlets in the microgrid, and a supply of generated electric power in the microgrid at a given time.

  19. MRI of thoracic outlet syndrome in children

    International Nuclear Information System (INIS)

    Chavhan, Govind B.; Batmanabane, Vaishnavi; Muthusami, Prakash; Towbin, Alexander J.; Borschel, Gregory H.

    2017-01-01

    Thoracic outlet syndrome is caused by compression of the neurovascular bundle as it passes from the upper thorax to the axilla. The neurovascular bundle can be compressed by bony structures such as the first rib, cervical ribs or bone tubercles, or from soft-tissue abnormalities like a fibrous band, muscle hypertrophy or space-occupying lesion. Thoracic outlet syndrome commonly affects young adults but can be seen in the pediatric age group, especially in older children. Diagnosis is based on a holistic approach encompassing clinical features, physical examination findings including those triggered by various maneuvers, electromyography, nerve conduction studies and imaging. Imaging is performed to confirm the diagnosis, exclude mimics and classify thoracic outlet syndrome into neurogenic, arterial, venous or mixed causes. MRI and MR angiography are useful in this process. A complete MRI examination for suspected thoracic outlet syndrome should include the assessment of anatomy and any abnormalities using routine sequences, vessel assessment with the arms in adduction by MR angiography and assessment of dynamic compression of vessels with abduction of the arms. The purpose of this paper is to describe the anatomy of the thoracic outlet, causes of thoracic outlet syndrome, the MR imaging techniques used in its diagnosis and the principles of image interpretation. (orig.)

  20. MRI of thoracic outlet syndrome in children

    Energy Technology Data Exchange (ETDEWEB)

    Chavhan, Govind B.; Batmanabane, Vaishnavi [The Hospital for Sick Children and University of Toronto, Department of Diagnostic Imaging, Toronto, ON (Canada); Muthusami, Prakash [The Hospital for Sick Children and University of Toronto, Department of Diagnostic Imaging, Toronto, ON (Canada); The Hospital for Sick Children, Division of Image Guided Therapy, Department of Diagnostic Imaging, Toronto, ON (Canada); Towbin, Alexander J. [Cincinnati Children' s Hospital Medical Center, Department of Radiology and Medical Imaging, Cincinnati, OH (United States); Borschel, Gregory H. [The Hospital for Sick Children and University of Toronto, Division of Plastic Surgery, Department of Pediatric Surgery, Toronto, ON (Canada)

    2017-09-15

    Thoracic outlet syndrome is caused by compression of the neurovascular bundle as it passes from the upper thorax to the axilla. The neurovascular bundle can be compressed by bony structures such as the first rib, cervical ribs or bone tubercles, or from soft-tissue abnormalities like a fibrous band, muscle hypertrophy or space-occupying lesion. Thoracic outlet syndrome commonly affects young adults but can be seen in the pediatric age group, especially in older children. Diagnosis is based on a holistic approach encompassing clinical features, physical examination findings including those triggered by various maneuvers, electromyography, nerve conduction studies and imaging. Imaging is performed to confirm the diagnosis, exclude mimics and classify thoracic outlet syndrome into neurogenic, arterial, venous or mixed causes. MRI and MR angiography are useful in this process. A complete MRI examination for suspected thoracic outlet syndrome should include the assessment of anatomy and any abnormalities using routine sequences, vessel assessment with the arms in adduction by MR angiography and assessment of dynamic compression of vessels with abduction of the arms. The purpose of this paper is to describe the anatomy of the thoracic outlet, causes of thoracic outlet syndrome, the MR imaging techniques used in its diagnosis and the principles of image interpretation. (orig.)

  1. Mapping of sodium void worth and doppler effect for sodium-cooled fast reactor - 15458

    International Nuclear Information System (INIS)

    Krepel, J.; Pelloni, S.; Bortot, S.; Panadero, A.L.; Mikityuk, K.

    2015-01-01

    The sodium-cooled fast reactor (SFR) represents the reference and the most technologically mastered system among the Generation-IV reactors. Nevertheless, the sodium void worth in the fuel regions of SFR is usually positive. To overcome this safety drawback, low-void sodium-cooled fast spectrum core (CFV) was proposed by CEA. Such a CFV core is used in the frame of WP6 'Core safety' of the FP7 Euratom ESNII+ project as a reference SFR design. The overall sodium void effect is negative for the CFV core. Nevertheless, locally it is positive in the fuel region and negative in the sodium plenum. Similarly, also the Doppler effect is spatially dependent and it varies between the inner and outer fuel regions and between the middle and lower blankets. Accordingly, knowledge of the local distributions or actually mappings of the two safety-related parameters will be necessary, before safety assessment and transient analysis can be done. In this study these maps have been produced using the deterministic code ERANOS. The obtained mapping shows strong local dependency of both safety-related effects. A sensitivity of the void effect to the sodium plenum modeling was also demonstrated. The results may serve as an input for the transient analysis of the CFV core or as a cross-check for the Monte Carlo method based maps. (authors)

  2. Alcohol Outlets and Violent Crime in Washington D.C.

    Directory of Open Access Journals (Sweden)

    Pan, William K

    2010-08-01

    Full Text Available Objective: Alcohol is more likely than any other drug to be involved in substance-related violence. In 2000 violence-related and self-directed injuries accounted for an estimated $37 billion and $33 billion in productivity losses and medical treatment, respectively. A review of emergency department data revealed violence and clinically identified trauma-related injuries have the strongest correlation among alcohol-dependent injuries. At the environmental level there is a relationship between alcohol outlet density and violent crime. A limited number of studies have examined the relationship between alcohol outlet type and the components of violent crime. The aim of this study is to examine the relationship between the aggregate components of violent crime and alcohol outlet density by type of outlet.Methods: For this study we used Washington, D.C. census tract data from the 2000 census to examine neighborhood characteristics. Alcohol outlet, violent crime, and population-level data for Washington, D.C. were drawn from various official yet publicly available sources. We developed an analytic database to examine the relationship between alcohol outlet category and four types of violent crime. After estimating spatial correlation and determining spatial dependence, we used a negative binomial regression analysis to assess the alcohol availability-violent crime association, while controlling for structural correlates of violence.Results: Independent of alternative structural correlates of violent crime, including the prevalence of weapons and illicit drugs, community-level alcohol outlet density is significantly associated with assaultive violence. Outlets were significantly related to robbery, assault, and sexual offenses. In addition, the relationship among on-premise and off-premise outlets varied across violent crime categories.Conclusion: In Washington, D.C., alcohol outlet density is significantly associated with the violent crimes. The

  3. Experimental investigations of heat transfer during sodium boiling in fuel assembly model in justification of advanced fast reactor safety

    International Nuclear Information System (INIS)

    Khafizov, R.R.; Poplavskij, V.M.; Rachkov, V.I.; Sorokin, A.P.; Ashurko, Yu.M.; Volkov, A.V.; Ivanov, E.F.; Privezentsev, V.V.

    2015-01-01

    The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m 2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling [ru

  4. 49 CFR 178.345-11 - Tank outlets.

    Science.gov (United States)

    2010-10-01

    ... unloading of lading, as distinguished from outlets such as manhole covers, vents, vapor recovery devices... away from the loading/unloading outlet. The actuating mechanism must be corrosion-resistant and...

  5. 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.

  6. Hydrologic Outlets of the Greenland Ice Sheet

    Data.gov (United States)

    National Aeronautics and Space Administration — The Hydrologic Outlets of the Greenland Ice Sheet data set contains GIS point shapefiles that include 891 observed and potential hydrologic outlets of the Greenland...

  7. Experience with reactor assembly of FBTR

    International Nuclear Information System (INIS)

    Srinivasan, G.; Ravishankar, K.; Babu, A.; Varadarajan, S.; Arumugam, P.; Sekhar, P.

    2006-01-01

    Reactor Assembly, also called Block Pile, is the heart of FBTR and houses the core, top and lateral shields, control rod drive mechanisms (CRDM), sodium inlet pipe and outlet pipes etc. Two major problems which arose during commissioning were reactor vessel tilt due to convection in cover gas space and failure of inflatable seals. The reactor vessel tilt was solved by Helium injection. Reactor was operated without pressurising the inflatable seals till 2005, when the seals were replaced. Other major problems in the course of twenty years of reactor operation were failure of three CRDM lower parts, Core Cover plate which houses the core thermocouples getting stuck in the fuel handling position, water leaks from the Biological Shield Cooling (BSC) coils around the reactor, failure of core wires in the trailing cables during fuel handling etc. This paper addresses the major problems faced and modifications carried out. (author)

  8. Parametric Evaluation of Large-Scale High-Temperature Electrolysis Hydrogen Production Using Different Advanced Nuclear Reactor Heat Sources

    International Nuclear Information System (INIS)

    Harvego, Edwin A.; McKellar, Michael G.; O'Brien, James E.; Herring, J. Stephen

    2009-01-01

    High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered

  9. 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)

  10. Validation and verification of the MTR{sub P}C thermohydraulic package

    Energy Technology Data Exchange (ETDEWEB)

    Doval, Alicia [INVAP S.E., Bariloche, Rio Negro (Argentina). Nuclear Engineering Dept.]. E-mail: doval@invap.com.ar

    1998-07-01

    The MTR{sub P}C v2.6 is a computational package developed for research reactor design and calculation. It covers three of the main aspects of a research reactor: neutronic, shielding and thermohydraulic. In this work only the thermohydraulic package will be covered, dealing with verification and validation aspects. The package consists of the following steady state programs: CAUDVAP 2.60 for the hydraulic calculus, estimates the velocity distribution through different parallel channels connected to a common inlet and outlet common plenum. TERMIC 1H v3.0, used for the thermal design of research reactors, provides information about heat flux for a given maximum wall temperature, onset of nucleate boiling, redistribution phenomena and departure from nucleate boiling. CONVEC V3.0 allows natural convection calculations, giving information on heat fluxes for onset of nucleate boiling, pulsed and burn-out phenomena as well as total coolant flow. Results have been validated against experimental values and verified against theoretical and computational programmes results, showing a good agreement. (author)

  11. Monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors

    International Nuclear Information System (INIS)

    Stanc, S.; Repa, M.

    2001-01-01

    Description of a monitoring system for accuracy and reliability characteristics of standard temperature measurements in WWER-440 reactors and benefits obtained from its use are shown in the presentation. As standard reactor temperature measurement, coolant temperature measurement at fuel assembly outlets and in loops, entered into the In-Reactor Control System , are considered. Such systems have been implemented at two V-230 reactors and are under implementation at other four V-213 reactors. (Authors)

  12. 1-D Two-phase Flow Investigation for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Kim, Jae Cheol

    2007-02-01

    During a severe accident, when a molten corium is relocated in a reactor vessel lower head, the RCF(Reactor Cavity Flooding) system for ERVC (External Reactor Vessel Cooling) is actuated and coolants are supplied into a reactor cavity to remove a decay heat from the molten corium. This severe accident mitigation strategy for maintaining a integrity of reactor vessel was adopted in the nuclear power plants of APR1400, AP600, and AP1000. Under the ERVC condition, the upward two-phase flow is driven by the amount of the decay heat from the molten corium. To achieve the ERVC strategy, the two-phase natural circulation in the annular gap between the external reactor vessel and the insulation should be formed sufficiently by designing the coolant inlet/outlet area and gap size adequately on the insulation device. Also the natural circulation flow restriction has to be minimized. In this reason, it is needed to review the fundamental structure of insulation. In the existing power plants, the insulation design is aimed at minimizing heat losses under a normal operation. Under the ERVC condition, however, the ability to form the two-phase natural circulation is uncertain. Namely, some important factors, such as the coolant inlet/outlet areas, flow restriction, and steam vent etc. in the flow channel, should be considered for ERVC design. T-HEMES 1D study is launched to estimate the natural circulation flow under the ERVC condition of APR1400. The experimental facility is one-dimensional and scaled down as the half height and 1/238 channel area of the APR1400 reactor vessel. The air injection method was used to simulate the boiling at the external reactor vessel and generate the natural circulation two-phase flow. From the experimental results, the natural circulation flow rate highly depended on inlet/outlet areas and the circulation flow rate increased as the outlet height as well as the supplied water head increased. On the other hand, the simple analysis using the drift

  13. The geography of Fast Food outlets: a review.

    Science.gov (United States)

    Fraser, Lorna K; Edwards, Kimberly L; Cade, Janet; Clarke, Graham P

    2010-05-01

    The availability of food high in fat, salt and sugar through Fast Food (FF) or takeaway outlets, is implicated in the causal pathway for the obesity epidemic. This review aims to summarise this body of research and highlight areas for future work. Thirty three studies were found that had assessed the geography of these outlets. Fourteen studies showed a positive association between availability of FF outlets and increasing deprivation. Another 13 studies also included overweight or obesity data and showed conflicting results between obesity/overweight and FF outlet availability. There is some evidence that FF availability is associated with lower fruit and vegetable intake. There is potential for land use policies to have an influence on the location of new FF outlets. Further research should incorporate good quality data on FF consumption, weight and physical activity.

  14. The Geography of Fast Food Outlets: A Review

    Directory of Open Access Journals (Sweden)

    Lorna K. Fraser

    2010-05-01

    Full Text Available The availability of food high in fat, salt and sugar through Fast Food (FF or takeaway outlets, is implicated in the causal pathway for the obesity epidemic. This review aims to summarise this body of research and highlight areas for future work. Thirty three studies were found that had assessed the geography of these outlets. Fourteen studies showed a positive association between availability of FF outlets and increasing deprivation. Another 13 studies also included overweight or obesity data and showed conflicting results between obesity/overweight and FF outlet availability. There is some evidence that FF availability is associated with lower fruit and vegetable intake. There is potential for land use policies to have an influence on the location of new FF outlets. Further research should incorporate good quality data on FF consumption, weight and physical activity.

  15. Tests of the RBMK-1500 reactor fuel assemblies in the Leningrad reactor

    International Nuclear Information System (INIS)

    Aden, V.C.; Varovin, I.A.; Vorontsov, B.A.

    1981-01-01

    Test of fuel assemblies of the RBMK-1500 reactor is conducted in the reactor of the Leningrad NPP unit 2 for proving the calculational values of critical power of the RBMK-1500 reactor fuel assemblies adopted in design. The experiment presupposes the maximal approximation of the fuel assembly operation parameters to the calculational critical parameters without bringing into the mode of heat transfer crisis. The experiments are carried out at 500, 850 and 900 MW(el) of the reactor. The maximal channel power made up 472 kW at 20.5 t/h coolant flow rate and 49% mass steam content at the outlet of the channel. It was concluded that there was supply up to the heat transfer crisis in all the investigated modes. Data of temperature measurings of the fuel element cans, readings of the devices of the failure control system of the fuel element cans and external inspection of the assemblies after the tests testify to it [ru

  16. Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

    2007-03-01

    The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

  17. Burnable poison rod for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Funk, C.E.; Oneufer, A.S.

    1984-01-01

    A burnable poison rod for use in a nuclear reactor fuel assembly which includes concentrically disposed rods having an annular space therebetween which extends the full length of the rods. The inner rod is hollow to permit circulation of coolant therethrough. Annular burnable poison pellets are positioned in the annular space which is closed at both ends by plugs. A spring clip is located in the plenum space above the pellet stack in the rods. The spring clip is of cylindrical configuration having a gap in the material which provides two ends adapted to be squeezed toward each other. A cross section of the clip shows that its ends contain alternating flat and round edges, the round edges conforming to the outer rod inner surface to provide a retentive force which is releasably applied to the pellet stack as it grows during operation in a reactor

  18. Reactor safety issues resolved by the 2D/3D program

    International Nuclear Information System (INIS)

    1995-09-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated. This report was prepared in a coordination among US, Germany and Japan. US and Germany have published the report as NUREG/IA-0127 and GRS-101 respectively. (author)

  19. Reactor safety issues resolved by the 2D/3D program

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    The 2D/3D Program studied multidimensional thermal-hydraulics in a PWR core and primary system during the end-of-blowdown and post-blowdown phases of a large-break LOCA (LBLOCA), and during selected small-break LOCA (SBLOCA) transients. The program included tests at the Cylindrical Core Test Facility (CCTF), the Slab Core Test Facility (SCTF), and the Upper Plenum Test Facility (UPTF), and computer analyses using TRAC. Tests at CCTF investigated core thermal-hydraulics and overall system behavior while tests at SCTF concentrated on multidimensional core thermal-hydraulics. The UPTF tests investigated two-phase flow behavior in the downcomer, upper plenum, tie plate region, and primary loops. TRAC analyses evaluated thermal-hydraulic behavior throughout the primary system in tests as well as in PWRs. This report summarizes the test and analysis results in each of the main areas where improved information was obtained in the 2D/3D Program. The discussion is organized in terms of the reactor safety issues investigated. This report was prepared in a coordination among US, Germany and Japan. US and Germany have published the report as NUREG/IA-0127 and GRS-101 respectively. (author).

  20. Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

    International Nuclear Information System (INIS)

    Zhang Longfei; Zhang Dafa; Wang Shaoming

    2007-01-01

    Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

  1. Tobacco Retail Outlets and Vulnerable Populations in Ontario, Canada

    Directory of Open Access Journals (Sweden)

    Michael O. Chaiton

    2013-12-01

    Full Text Available Interest has been increasing in regulating the location and number of tobacco vendors as part of a comprehensive tobacco control program. The objective of this paper is to examine the distribution of tobacco outlets in a large jurisdiction, to assess: (1 whether tobacco outlets are more likely to be located in vulnerable areas; and (2 what proportion of tobacco outlets are located close to schools. Retail locations across the Province of Ontario from Ministry of Health Promotion data were linked to 2006 Census data at the neighbourhood level. There was one tobacco retail outlet for every 1,000 people over age 15 in Ontario. Density of outlets varied by public health unit, and was associated with the number of smokers. Tobacco outlets were more likely to be located in areas that had high neighbourhood deprivation, in both rural and urban areas. Outlets were less likely to be located in areas with high immigrant populations in urban areas, with the reverse being true for rural areas. Overall, 65% of tobacco retailers were located within 500 m of a school. The sale of tobacco products is ubiquitous, however, neighbourhoods with lower socio-economic status are more likely to have easier availability of tobacco products and most retailers are located within walking distance of a school. The results suggest the importance of policies to regulate the location of tobacco retail outlets.

  2. An evaluation of reactor cooling and coupled hydrogen production processes using the modular helium reactor

    International Nuclear Information System (INIS)

    Harvego, E.A.; Reza, S.M.M.; Richards, M.; Shenoy, A.

    2006-01-01

    The high-temperature characteristics of the modular helium reactor (MHR) make it a strong candidate for producing hydrogen using either thermochemical or high-temperature electrolysis (HTE) processes. Using heat from the MHR to drive a sulfur-iodine (SI) thermochemical hydrogen production process has been the subject of a U.S. Department of Energy sponsored Nuclear Engineering Research Initiative (NERI) project led by General Atomics, with participation from the Idaho National Laboratory (INL) and Texas A and M University. While the focus of much of the initial work was on the SI thermochemical production of hydrogen, recent activities included development of a preconceptual design for an integral HTE hydrogen production plant driven by the process heat and electricity produced by a 600 MW MHR. This paper describes ATHENA analyses performed to evaluate alternative primary system cooling configurations for the MHR to minimize peak reactor vessel and core temperatures while achieving core helium outlet temperatures in the range of 900-1000 deg. C that are needed for the efficient production of hydrogen using either the SI or HTE process. The cooling schemes investigated are intended to ensure peak fuel temperatures do not exceed specified limits under normal or transient upset conditions, and that reactor vessel temperatures do not exceed American Society of Mechanical Engineers (ASME) code limits for steady-state or transient conditions using standard light water reactor vessel materials. Preconceptual designs for SI and HTE hydrogen production plants driven by one or more 600 MW MHRs at helium outlet temperatures in the range of 900-1000 deg. C are described and compared. An initial SAPHIRE model to evaluate the reliability, maintainability, and availability of the SI hydrogen production plant is also described. Finally, a preliminary flowsheet for a conceptual design of an HTE hydrogen production plant coupled to a 600 MW modular helium reactor is presented and

  3. The Integration Of Process Heat Applications To High Temperature Gas Reactors

    International Nuclear Information System (INIS)

    McKellar, Michael G.

    2011-01-01

    A high temperature gas reactor, HTGR, can produce industrial process steam, high-temperature heat-transfer gases, and/or electricity. In conventional industrial processes, these products are generated by the combustion of fossil fuels such as coal and natural gas, resulting in significant emissions of greenhouse gases such as carbon dioxide. Heat or electricity produced in an HTGR could be used to supply process heat or electricity to conventional processes without generating any greenhouse gases. Process heat from a reactor needs to be transported by a gas to the industrial process. Two such gases were considered in this study: helium and steam. For this analysis, it was assumed that steam was delivered at 17 MPa and 540 C and helium was delivered at 7 MPa and at a variety of temperatures. The temperature of the gas returning from the industrial process and going to the HTGR must be within certain temperature ranges to maintain the correct reactor inlet temperature for a particular reactor outlet temperature. The returning gas may be below the reactor inlet temperature, ROT, but not above. The optimal return temperature produces the maximum process heat gas flow rate. For steam, the delivered pressure sets an optimal reactor outlet temperature based on the condensation temperature of the steam. ROTs greater than 769.7 C produce no additional advantage for the production of steam.

  4. Parameter analysis calculation on characteristics of portable FAST reactor

    International Nuclear Information System (INIS)

    Otsubo, Akira; Kowata, Yasuki

    1998-06-01

    In this report, we performed a parameter survey analysis by using the analysis program code STEDFAST (Space, TErrestrial and Deep sea FAST reactor-gas turbine system). Concerning the deep sea fast reactor-gas turbine system, calculations with many variable parameters were performed on the base case of a NaK cooled reactor of 40 kWe. We aimed at total equipment weight and surface area necessary to remove heat from the system as important values of the characteristics of the system. Electric generation power and the material of a pressure hull were specially influential for the weight. The electric generation power, reactor outlet/inlet temperatures, a natural convection heat transfer coefficient of sea water were specially influential for the area. Concerning the space reactor-gas turbine system, the calculations with the variable parameters of compressor inlet temperature, reactor outlet/inlet temperatures and turbine inlet pressure were performed on the base case of a Na cooled reactor of 40 kWe. The first and the second variable parameters were influential for the total equipment weight of the important characteristic of the system. Concerning the terrestrial fast reactor-gas turbine system, the calculations with the variable parameters of heat transferred pipe number in a heat exchanger to produce hot water of 100degC for cogeneration, compressor stage number and the kind of primary coolant material were performed on the base case of a Pb cooled reactor of 100 MWt. In the comparison of calculational results for Pb and Na of primary coolant material, the primary coolant weight flow rate was naturally large for the former case compared with for the latter case because density is very different between them. (J.P.N.)

  5. Predicted Variations of Water Chemistry in the Primary Coolant Circuit of a Supercritical Water Reactor

    International Nuclear Information System (INIS)

    Yeh, Tsung-Kuang; Wang, Mei-Ya; Liu, Hong-Ming; Lee, Min

    2012-09-01

    In response to the demand over a higher efficiency for a nuclear power plant, various types of Generation IV nuclear reactors have been proposed. One of the new generation reactors adopts supercritical light water as the reactor coolant. While current in-service light water reactors (LWRs) bear an average thermal efficiency of 33%, the thermal efficiency of a supercritical water reactor (SCWR) could generally reach more than 44%. For LWRs, the coolants are oxidizing due to the presence of hydrogen peroxide and oxygen, and the degradation of structural materials has mainly resulted from stress corrosion cracking. Since oxygen is completely soluble in supercritical water, similar or even worse degradation phenomena are expected to appear in the structural and core components of an SCWR. To ensure proper designs of the structural components and suitable selections of the materials to meet the requirements of operation safety, it would be of great importance for the design engineers of an SCWR to be fully aware of the state of water chemistry in the primary coolant circuit (PCC). Since SCWRs are still in the stage of conceptual design and no practical data are available, a computer model was therefore developed for analyzing water chemistry variation and corrosion behavior of metallic materials in the PCC of a conceptual SCWR. In this study, a U.S. designed SCWR with a rated thermal power of 3575 MW and a coolant flow rate of 1843 kg/s was selected for investigating the variations in redox species concentration in the PCC. Our analyses indicated that the [H 2 ] and [H 2 O 2 ] at the core channel were higher than those at the other regions in the PCC of this SCWR. Due to the self-decomposition of H 2 O 2 , the core channel exhibited a lower [O 2 ] than the upper plenum. Because the middle water rod region was in parallel with the core channel region with relatively high dose rates, the [H 2 ] and [H 2 O 2 ] in this region were higher than those in the other regions

  6. 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

  7. Development of the temperature field at the WWER-440 core outlet monitoring system and application of the data analyses methods

    International Nuclear Information System (INIS)

    Spasova, V.; Georgieva, N.; Haralampieva, Tz.

    2001-01-01

    On-line internal reactor monitoring by 216 thermal couples, located at the reactor core outlet, is carried out during power operation of WWER-440 Units 1 and 2 at Kozloduy NPP. Automatic monitoring of technology process is performed by IB-500MA, which collects and performs initial data processing (discrediting and conversion of analogue signals into digital mode). The paper also presents the results and analyses of power distribution monitoring during the past 21-th and current 22-th fuel cycle at Kozloduy NPP, Unit 1 by using archiving system capacity and related software. The possibility to perform operational assessment and analysis of power distribution in the reactor core in each point of the fuel cycle is checked by comparison of the neutron-physical calculation results with reactor coolant system parameters. Paper shows that the processing and analysis of accumulated significant amount of data in the archive files increases accuracy and reliability of power distribution monitoring in the reactor core in each moment of the fuel cycle of WWER-440 reactors at Kozloduy NPP

  8. Fully coupled modeling of burnup dependent light water reactor fuel performance using COMSOL Multiphysics

    International Nuclear Information System (INIS)

    Liu Rong; Zhou Wenzhong; Prudil, Andrew

    2015-01-01

    This paper presents the development of a light water reactor fuel performance code, which considers almost all the related physical models, including heat generation and conduction, species diffusion, thermomechanics (thermal expansion, elastic strain, densification, and fission product swelling strain), grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, cladding thermal and irradiation creep and oxidation. All the equations are implemented into COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet and cladding. Comparisons are made for the simulation results between COMSOL and another simulation tool of BISON. The comparisons show the capability of our simulation tool to predict light water UO 2 fuel performances. In our modeling and simulation work, the performance of enhanced thermal conductivity UO 2 -BeO fuel and newly-adopted corrosion resistant SiC cladding material was also studied. UO 2 -BeO high thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction through lessening thermal stresses that result in fuel cracking, relocation, and swelling. The safety of the reactor would be improved. However, for SiC cladding, although due to its high thermal expansion, the gap closure time is delayed, irradiation induced point defects and defect-clusters in the SiC crystal will dramatically decrease SiC thermal conductivity, and cause significant increase in the fuel temperature. (author)

  9. Assessing the degree of plug flow in oxidation flow reactors (OFRs: a study on a potential aerosol mass (PAM reactor

    Directory of Open Access Journals (Sweden)

    D. Mitroo

    2018-03-01

    Full Text Available Oxidation flow reactors (OFRs have been developed to achieve high degrees of oxidant exposures over relatively short space times (defined as the ratio of reactor volume to the volumetric flow rate. While, due to their increased use, attention has been paid to their ability to replicate realistic tropospheric reactions by modeling the chemistry inside the reactor, there is a desire to customize flow patterns. This work demonstrates the importance of decoupling tracer signal of the reactor from that of the tubing when experimentally obtaining these flow patterns. We modeled the residence time distributions (RTDs inside the Washington University Potential Aerosol Mass (WU-PAM reactor, an OFR, for a simple set of configurations by applying the tank-in-series (TIS model, a one-parameter model, to a deconvolution algorithm. The value of the parameter, N, is close to unity for every case except one having the highest space time. Combined, the results suggest that volumetric flow rate affects mixing patterns more than use of our internals. We selected results from the simplest case, at 78 s space time with one inlet and one outlet, absent of baffles and spargers, and compared the experimental F curve to that of a computational fluid dynamics (CFD simulation. The F curves, which represent the cumulative time spent in the reactor by flowing material, match reasonably well. We value that the use of a small aspect ratio reactor such as the WU-PAM reduces wall interactions; however sudden apertures introduce disturbances in the flow, and suggest applying the methodology of tracer testing described in this work to investigate RTDs in OFRs to observe the effect of modified inlets, outlets and use of internals prior to application (e.g., field deployment vs. laboratory study.

  10. Assessing the degree of plug flow in oxidation flow reactors (OFRs): a study on a potential aerosol mass (PAM) reactor

    Science.gov (United States)

    Mitroo, Dhruv; Sun, Yujian; Combest, Daniel P.; Kumar, Purushottam; Williams, Brent J.

    2018-03-01

    Oxidation flow reactors (OFRs) have been developed to achieve high degrees of oxidant exposures over relatively short space times (defined as the ratio of reactor volume to the volumetric flow rate). While, due to their increased use, attention has been paid to their ability to replicate realistic tropospheric reactions by modeling the chemistry inside the reactor, there is a desire to customize flow patterns. This work demonstrates the importance of decoupling tracer signal of the reactor from that of the tubing when experimentally obtaining these flow patterns. We modeled the residence time distributions (RTDs) inside the Washington University Potential Aerosol Mass (WU-PAM) reactor, an OFR, for a simple set of configurations by applying the tank-in-series (TIS) model, a one-parameter model, to a deconvolution algorithm. The value of the parameter, N, is close to unity for every case except one having the highest space time. Combined, the results suggest that volumetric flow rate affects mixing patterns more than use of our internals. We selected results from the simplest case, at 78 s space time with one inlet and one outlet, absent of baffles and spargers, and compared the experimental F curve to that of a computational fluid dynamics (CFD) simulation. The F curves, which represent the cumulative time spent in the reactor by flowing material, match reasonably well. We value that the use of a small aspect ratio reactor such as the WU-PAM reduces wall interactions; however sudden apertures introduce disturbances in the flow, and suggest applying the methodology of tracer testing described in this work to investigate RTDs in OFRs to observe the effect of modified inlets, outlets and use of internals prior to application (e.g., field deployment vs. laboratory study).

  11. Preliminary Sensitivity Study on Gas-Cooled Reactor for NHDD System Using MARS-GCR

    International Nuclear Information System (INIS)

    Lee, Seung Wook; Jeong, Jae Jun; Lee, Won Jae

    2005-01-01

    A Gas-Cooled Reactor (GCR) is considered as one of the most outstanding tools for a massive hydrogen production without CO 2 emission. Till now, two types of GCR are regarded as a viable nuclear reactor for a hydrogen production: Prismatic Modular Reactor (PMR), Pebble Bed Reactor (PBR). In this paper, a preliminary sensitivity study on two types of GCR is carried out by using MARS-GCR to find out the effect on the peak fuel and reactor pressure vessel (RPV) temperature, with varying the condition of a reactor inlet, outlet temperature, and system pressure for both PMR and PBR

  12. Fully coupled multiphysics modeling of enhanced thermal conductivity UO{sub 2}–BeO fuel performance in a light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, R. [Department of Mechanical and Biomedical Engineering, City University of Hong Kong, Hong Kong (China); Zhou, W., E-mail: wenzzhou@cityu.edu.hk [Department of Mechanical and Biomedical Engineering, City University of Hong Kong, Hong Kong (China); Shen, P. [Department of Mechanical and Biomedical Engineering, City University of Hong Kong, Hong Kong (China); Prudil, A. [Fuel and Fuel Channel Safety Branch, Canadian Nuclear Laboratories, Chalk River, Ontario (Canada); Chan, P.K. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, Ontario (Canada)

    2015-12-15

    Highlights: • LWR fuel performance modeling capability developed. • Fully coupled multiphysics studies for enhanced thermal conductivity UO{sub 2}–BeO fuel. • UO{sub 2}–BeO fuel decreases fuel temperature and lessens thermal stresses. • UO{sub 2}–BeO fuel facilitates a reduction in PCMI. • Reactor safety can be improved for UO{sub 2}–BeO fuel. - Abstract: Commercial light water reactor fuel UO{sub 2} has a low thermal conductivity that leads to the development of a large temperature gradient across the fuel pellet, limiting the reactor operational performance due to the effects that include thermal stresses causing pellet cladding interaction and the release of fission product gases. This study presents the development of a modeling and simulation for enhanced thermal conductivity UO{sub 2}–BeO fuel behavior in a light water reactor, using self-defined multiple physics models fully coupled based on the framework of COMSOL Multiphysics. Almost all the related physical models are considered, including heat generation and conduction, species diffusion, thermomechanics (thermal expansion, elastic strain, densification, and fission product swelling strain), grain growth, fission gas production and release, gap heat transfer, mechanical contact, gap/plenum pressure with plenum volume, cladding thermal and irradiation creep and oxidation. All the phenomenal models and materials properties are implemented into COMSOL Multiphysics finite-element platform with a 2D axisymmetric geometry of a fuel pellet and cladding. UO{sub 2}–BeO enhanced thermal conductivity nuclear fuel would decrease fuel temperatures and facilitate a reduction in pellet cladding interaction from our simulation results through lessening thermal stresses that result in fuel cracking, relocation, and swelling, so that the safety of the reactor would be improved.

  13. Study on the transient behaviours of MNSR reactor for control rod withdrawal

    International Nuclear Information System (INIS)

    Yang Shunhai

    1995-10-01

    The transient behaviours of Miniature Neutron Source Reactor MNSR are analyzed and calculated with the reactor thermohydraulics RETRAN-02 program and the reactor physics MARIA program. The obtained event sequence and consequence from the calculation are compared with the experiments. The effective resonance integral for study on Doppler effect is taken into account. The reactivity temperature coefficient weighting factors are computed. The transient parameters related to reactor power peaking, coolant inlet temperatures, outlet temperatures and coolant mass flow, etc. are computed and compared with the experimental results. (6 refs., 2 figs., 5 tabs.)

  14. Nuclear fuel element containing strips of an alloyed Zr, Ti, and Ni getter material

    International Nuclear Information System (INIS)

    Grossman, L.N.; Packard, D.R.

    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 strips and preferably the strips are positioned inside a helical member in the plenum. The position of the alloy strips permits gases and liquids entering the plenum to contact and react with the alloy strips. (U.S.)

  15. Ceramic membrane reactor with two reactant gases at different pressures

    Science.gov (United States)

    Balachandran, Uthamalingam; Mieville, Rodney L.

    2001-01-01

    The invention is a ceramic membrane reactor for syngas production having a reaction chamber, an inlet in the reactor for natural gas intake, a plurality of oxygen permeating ceramic slabs inside the reaction chamber with each slab having a plurality of passages paralleling the gas flow for transporting air through the reaction chamber, a manifold affixed to one end of the reaction chamber for intake of air connected to the slabs, a second manifold affixed to the reactor for removing the oxygen depleted air, and an outlet in the reaction chamber for removing syngas.

  16. Simulation of emulsion copolymerization reactions in a continuous pulsed sieve-plate column reactor

    OpenAIRE

    C. Sayer; R. Giudici

    2004-01-01

    This work addressed the viability of using a pulsed sieve-plate column reactor to carry out continuous vinyl acetate/butyl acrylate emulsion copolymerization reactions. A rigorous mathematical model of emulsion copolymerization reactions in a tubular reactor with axial dispersion was used for this purpose. Operational conditions were defined to attain high monomer conversions at the reactor outlet in a relatively short residence time and, at the same time, produce a copolymer with a more homo...

  17. Analysis of the Sales Promotion in Choice Retail Outlet

    OpenAIRE

    HUMPOLCOVÁ, Michaela

    2010-01-01

    My bachelor thesis is aimed at sales promotion in a retail outlet. The main aim of this thesis is evaluate the current state of sales promotion in a selected retail outlet and based on the analysis of the current state of sales promotion in the outlet to try to propose some measures of improve.

  18. Performance of metallic fuels in liquid-metal fast reactors

    International Nuclear Information System (INIS)

    Seidel, B.R.; Walters, L.C.; Kittel, J.H.

    1984-01-01

    Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

  19. Hydrogen generation using the modular helium reactor

    International Nuclear Information System (INIS)

    Richards, M.; Shenoy, A.

    2004-01-01

    Process heat from a high-temperature nuclear reactor can be used to drive a set of chemical reactions, with the net result of splitting water into hydrogen and oxygen. For example, process heat at temperatures in the range 850 deg.C to 950 deg.C can drive the sulfur-iodine (SI) thermochemical process to produce hydrogen with high efficiency. Electricity can also be used to split water, using conventional, low-temperature electrolysis. An example of a hybrid process is high-temperature electrolysis (HTE), in which process heat is used to generate steam, which is then supplied to an electrolyser to generate hydrogen. In this paper we investigate the coupling of the Modular Helium Reactor (MHR) to the SI process and HTE. These concepts are referred to as the H2-MHR. Optimization of the MHR core design to produce higher coolant outlet temperatures is also discussed. The use of fixed orifices to control the flow distribution is a promising design solution for increasing the coolant outlet temperature without increasing peak fuel temperatures significantly

  20. Core of a liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Wright, J.R.; McFall, A.

    1975-01-01

    The core of a liquid-cooled nuclear reactor, e.g. of a sodium-cooled fast reactor, is protected in such a way that the recoil wave resulting from loss of coolant in a cooling channel and caused by released gas is limited to a coolant inlet chamber of this cooling channel. The channels essentially consist of the coolant inlet chamber and a fuel chamber - with a fission gas storage plenum - through which the coolant flows. Between the two chambers, a locking device within a tube is provided offering a much larger flow resistance to the backflow of gas or coolant than in flow direction. The locking device may be a hydraulic countertorque control system, e.g. a valvular line. Other locking devices have got radially helical vanes running around an annular flow space. Furthermore, the locking device may consist of a number of needles running parallel to each other and forming a circular grid. Though it can be expanded by the forward flow - the needles are spreading - , it acts as a solid barrier for backflows. (TK) [de

  1. Numerical predictions of natural convection in a uniformly heated pool

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Cho, D.H.

    1993-01-01

    In the event of a core meltdown accident, one of the accident progression paths is fuel relocation to the lower reactor plenum. In the heavy-water new production reactor (NPR-HWR) design, the reactor cavity is flooded with water. In such a design, decay heat removal to the water in the reactor cavity and thence to the containment may be adequate to keep the reactor vessel temperature below failure limits. If this is the case, the accident progression can be arrested by retaining a coolable corium configuration in the lower reactor plenum. The strategy of reactor cavity flooding to prevent reactor vessel failure from molten corium relocation to the reactor vessel lower head has also been considered for commercial pressurized water reactors

  2. Availability of limited service food outlets surrounding schools in British Columbia.

    Science.gov (United States)

    Black, Jennifer L; Day, Meghan

    2012-06-05

    The purpose of this study was to provide a descriptive profile of the availability of limited service food outlets surrounding public schools in British Columbia, Canada. Data from the 2010 Canadian Business Data Files were used to identify limited service food outlets including fast food outlets, beverage and snack food stores, delis and convenience stores. The number of food outlets within 800 metres of 1,392 public schools and the distance from schools to the nearest food outlets were assessed. Multivariate regression models examined the associations between food outlet availability and school-level characteristics. In 2010, over half of the public schools in BC (54%) were located within a 10-12 minute walk from at least one limited service food outlet. The median closest distance to a food outlet was just over 1 km (1016 m). Schools comprised of students living in densely populated urban neighbourhoods and neighbourhoods characterized by lower socio-economic status were more likely to have access to limited service food outlets within walking distance. After adjusting for school-level median family income and population density, larger schools had higher odds of exposure to food vendors compared to schools with fewer students. The availability of and proximity to limited service food outlets vary widely across schools in British Columbia and school-level characteristics are significantly associated with food outlet availability. Additional research is needed to understand how food environment exposures inside and surrounding schools impact students' attitudes, food choices and dietary quality.

  3. Thoracic outlet syndrome: Case report

    International Nuclear Information System (INIS)

    Marquez, Juan Camilo; Acosta, Mauricio Fernando; Uribe Jorge Ricardo

    2009-01-01

    We report a case of vascular thoracic outlet syndrome in a young man, diagnosed with upper limb arteriography, leading to repeated arterio-arterial emboli originating from a post-stenotic subclavian artery aneurysm. It is of our interest due to its low incidence and the small number of cases reported that have been diagnosed by arteriography. The thoracic outlet is the path through which vascular and neural structures goes from the neck to the axilla, and it has three anatomical strictures, that when pronounced, can compress the brachial plexus or subclavian vessels, leading to different symptoms and signs.

  4. Analytical evaluation of two-phase natural circulation flow characteristics under external reactor vessel cooling

    International Nuclear Information System (INIS)

    Park, Jong Woon

    2009-01-01

    This work proposes an analytical method of evaluating the effects of design and operating parameters on the low-pressure two-phase natural circulation flow through the annular shaped gap at the reactor vessel exterior surface heated by corium (molten core) relocated to the reactor vessel lower plenum after loss of coolant accidents. A natural circulation flow velocity equation derived from steady-state mass, momentum, and energy conservation equations for homogeneous two-phase flow is numerically solved for the core melting conditions of the APR1400 reactor. The solution is compared with existing experiments which measured natural circulation flow through the annular gap slice model. Two kinds of parameters are considered for this analytical method. One is the thermal-hydraulic conditions such as thermal power of corium, pressure and inlet subcooling. The others are those for the thermal insulation system design for the purpose of providing natural circulation flow path outside the reactor vessel: inlet flow area, annular gap clearance and system resistance. A computer program NCIRC is developed for the numerical solution of the implicit flow velocity equation.

  5. FBR and RBR particle bed space reactors

    International Nuclear Information System (INIS)

    Powell, J.R.; Botts, T.E.

    1983-01-01

    Compact, high-performance nuclear reactor designs based on High-Temperature Gas Reactors (HTGRs) particulate fuel are investigated. The large surface area available with the small-diameter (approx. 500 microns) particulate fuel allows very high power densities (MW's/liter), small temperature differences between fuel and coolant (approx. 10 0 K), high coolant-outlet temperatures (1500 to 3000 0 K, depending on design), and fast reactor startup (approx. 2 to 3 seconds). Two reactor concepts are developed - the Fixed Bed Reactor (FBR), where the fuel particles are packed into a thin annular bed between two porous cylindrical drums, and the Rotating Bed Reactor (RBR), where the fuel particles are held inside a cold rotating (typically approx. 500 rpm) porous cylindrical drum. The FBR can operate steady-state in the closed-cycle He-cooled mode or in the open-cycle H 2 -cooled mode. The RBR will operate only in the open-cycle H 2 -cooled mode

  6. C-Reactor I and E loading instability limits

    Energy Technology Data Exchange (ETDEWEB)

    Hess, K.W.

    1957-01-24

    The pilot charging of I & E fuel elements has been implemented at C-Reactor under Production Test IP-19-A. It was necessary to provide adequate tube protection against flow interruption by establishing proper trip setting on the Panellit pressure gauges. the administration of these Panellit trip settings is done by trip-before- boiling tube outlet temperature limits, which are similar in principle to the current instability limits. Trip-before-boiling limits for C-Reactor I & E fuel elements loadings are presented in this document.

  7. Design of the upper internals structure for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Thompson, D.C.; Novendstern, E.H.

    1977-01-01

    The Upper Internals Structure (UIS) is located above the core and is supported from the head at four locations. It is designed to perform the following primary functions: provide secondary core holddown in the event of a malfunction of the core hydraulic holddown system; provide support for routing all in-vessel instrumentation to core assemblies; maintain alignment between the core assemblies, the UIS and the closure head; provide guidance and crossflow protection for the control rod drivelines; and mix/duct flow to the upper region of the vessel outlet plenum to minimize rapid temperature changes to components during a reactor trip transient. In accomplishing these functions, the UIS will experience a sodium environment with temperatures up to 1200 0 F (649 0 C), and as many as 7 x 10 8 cycles of fluid temperature fluctuations up to 250 0 F (121 0 C) at full power operation. It must be designed to survive these conditions in combination with seismic and flow-induced vibration loadings for its 30 year design life. The design program of designing to controlled functional requirements and design conditions is discussed. Included is a description of the significant parts of the design and the approach used to balance the requirement of tight joints. The thermal and hydraulic environment including the results of a comprehensive test program are discussed. The test program results establish the basis of the thermal boundary used in the structural evaluation, and the UIS vibration characteristics. A summary of the areas which have required design changes is included with a summary of the structural evaluation of these changes

  8. 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.

  9. Numerical simulation for debris bed behavior in sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Tagami, Hirotaka; Tobita, Yoshiharu

    2014-01-01

    For safety analysis of SFR, it is necessary to evaluate behavior along with coolability of debris bed in lower plenum which is formed in severe accident. In order to analyze debris behavior, model for dense sediment particles behavior was proposed and installed in SFR safety analysis code SIMMER. SIMMER code could adequately reproduce experimental results simulating the self-leveling phenomena with appropriate model parameters for bed stiffness. In reactor condition, the self-leveling experiment for prototypical debris bed has not been performed. Additionally, the prototypical debris bed consists of non-spherical particles and it is difficult to quantify model parameters. This situation brings sensitivity analysis to investigate effect of model parameters on the self-leveling phenomena of prototypical debris bed in present paper. As initial condition for sensitivity analysis, simple mound-like debris bed in sodium-filled lower plenum in reactor vessel is considered. The bed consists of the mixture of fuel debris of 3,300 kg and steel debris of 1,570 kg. Decay heat is given to this fuel debris. The model parameter is chosen as sensitivity parameter. Sensitivity analysis shows that the model parameters can effect on intensity of self-leveling phenomena and eventual flatness of bed. In all analyses, however, coolant and sodium vapor break the debris bed at mainly center part of bed and the debris is relocated to outside of bed. Through this process, the initial debris bed is almost planarized before re-melting of debris. This result shows that the model parameters affect the self-leveling phenomena, but its effect in the safety analysis of SFRs is limited. (author)

  10. Simulation of emulsion copolymerization reactions in a continuous pulsed sieve-plate column reactor

    Directory of Open Access Journals (Sweden)

    Sayer C.

    2004-01-01

    Full Text Available This work addressed the viability of using a pulsed sieve-plate column reactor to carry out continuous vinyl acetate/butyl acrylate emulsion copolymerization reactions. A rigorous mathematical model of emulsion copolymerization reactions in a tubular reactor with axial dispersion was used for this purpose. Operational conditions were defined to attain high monomer conversions at the reactor outlet in a relatively short residence time and, at the same time, produce a copolymer with a more homogeneous composition.

  11. Fast reactor core monitoring device

    International Nuclear Information System (INIS)

    Sanda, Toshio; Inoue, Kotaro; Azekura, Kazuo.

    1982-01-01

    Purpose: To enable the rapid and accurate on-line identification of the state of a fast reactor core by effectively utilizing the measured data on the temperature and flow rate of the coolant. Constitution: The spacial power distribution and average assembly power are quickly calculated using an approximate calculating method, the measured values and the calculated values of the inlet and outlet temperature difference, flow rate and coolant physical values of an assembly are combined and are individually obtained, the most definite respective values and their errors are obtained by a least square method utilizing a formula of the relation between these values, and the power distribution and the temperature distribution of a reactor core are estimated in this manner. Accordingly, even when the measuring accuracy and the calculating accuracy are equal as in a fast reactor, the power distribution and the temperature distribution can be accurately estimated on-line at a high speed in a nuclear reactor, information required for the operator is provided, and the reactor can thus be safely and efficiently operated. (Yoshihara, H.)

  12. Conceptual design study of small lead-bismuth cooled reactor

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka; Hori, Toru; Kida, Masanori; Konomura, Mamoru

    2004-11-01

    In phase 2 of the feasibility study of commercialized fast reactor cycle systems of JNC, we make a concept of a small sodium cooled reactor for a power source of a city with various requirements, such as, safety and economical competitiveness. various reactor concepts are surveyed and a tank type reactor whose intermediate heat exchanger and primary main pumps are arranged in series is selected. In this study, a compact long life core and a simple reactor structure designs are pursued. The core type is three regional Zr concentration with one Pu enrichment core, the reactor outlet temperature achieves 550degC and the reactor electric output increases from 150 MWe to 165 MWe. The construction cost is much higher than the economical goal in the case of FOAK. But the construction cost in the case of NOAK is estimated to be 85.6% achieving the economical goal. (author)

  13. Control of reactor coolant flow path during reactor decay heat removal

    Science.gov (United States)

    Hunsbedt, Anstein N.

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  14. Research and development on next generation reactor (phase I)

    International Nuclear Information System (INIS)

    Park, Jong Kyoon; Chang, Moon Heuy; Hwang, Yung Dong

    1994-10-01

    The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author)

  15. Research and development on next generation reactor (phase I)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyoon; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); and others

    1994-10-01

    The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author).

  16. Studies on high temperature research reactor in China

    Energy Technology Data Exchange (ETDEWEB)

    Xu Yuanhui; Zuo Kanfen [Institute of Nuclear Energy Technology, Tsinghua Univ., Beijing (China)

    1999-08-01

    China recognises the advantages of Modular HTGRs and has chosen Modular HTGRs as one of advanced reactors to be developed for the further intensive utilisation of nuclear power in the next century. In energy supply systems of the next century, HTGR is supposed to serve: 1. as supplement to water-cooled reactors for electricity generation and 2. as environmentally friendly heat source providing process heat at different temperatures for various applications like heavy oil recovery, coal gasification and liquefaction, etc.. The 10 MW High Temperature Gas-cooled Reactor (HTR-10) is a major project in the energy sector of the Chinese National High Technology Programme as the first step of development of Modular HTGRs in China. Its main objectives are: 1. to acquire know-how in the design, construction and operation of HTGRs, 2. to establish an irradiation and experimental facility, 3. to demonstrate the inherent safety features of Modular HTGR, 4. to test electricity and heat co-generation and closed cycle gas turbine technology and 5. to do research and development work on the nuclear process heat application. Now the HTR-10 is being constructed at the site of Institute of Nuclear Energy Technology (INET). The HTR-10 project is to be carried out in two phases. In the first phase, the reactor with an coolant outlet temperature of 700degC will be coupled with a steam generator providing steam for a steam turbine cycle which works on an electricity and heat co-generation basis. In the second phase, the reactor coolant outlet temperature is planned to be raised to 900degC. As gas turbine cycle and a steam reformer will be coupled to the reactor in addition to the steam turbine cycle. (author)

  17. Studies on high temperature research reactor in China

    International Nuclear Information System (INIS)

    Xu Yuanhui; Zuo Kanfen

    1999-01-01

    China recognises the advantages of Modular HTGRs and has chosen Modular HTGRs as one of advanced reactors to be developed for the further intensive utilisation of nuclear power in the next century. In energy supply systems of the next century, HTGR is supposed to serve: 1. as supplement to water-cooled reactors for electricity generation and 2. as environmentally friendly heat source providing process heat at different temperatures for various applications like heavy oil recovery, coal gasification and liquefaction, etc.. The 10 MW High Temperature Gas-cooled Reactor (HTR-10) is a major project in the energy sector of the Chinese National High Technology Programme as the first step of development of Modular HTGRs in China. Its main objectives are: 1. to acquire know-how in the design, construction and operation of HTGRs, 2. to establish an irradiation and experimental facility, 3. to demonstrate the inherent safety features of Modular HTGR, 4. to test electricity and heat co-generation and closed cycle gas turbine technology and 5. to do research and development work on the nuclear process heat application. Now the HTR-10 is being constructed at the site of Institute of Nuclear Energy Technology (INET). The HTR-10 project is to be carried out in two phases. In the first phase, the reactor with an coolant outlet temperature of 700degC will be coupled with a steam generator providing steam for a steam turbine cycle which works on an electricity and heat co-generation basis. In the second phase, the reactor coolant outlet temperature is planned to be raised to 900degC. As gas turbine cycle and a steam reformer will be coupled to the reactor in addition to the steam turbine cycle. (author)

  18. Power distribution forecasting device for reactors

    International Nuclear Information System (INIS)

    Tsukii, Makoto

    1981-01-01

    Purpose: To save expensive calculations on the forecasting of reactor power distribution. Constitution: Core status (CSD) such as entire coolant flow rate, pressures in the reactor, temperatures at the outlet and inlet and positions for control rods are inputted into a power distribution calculation device to calculate the power distribution based on physical models intermittently. Further, present power distribution is calculated based on in-core neutron flux measured values and CSD in a process control computer. Further, the ratio of the calculation results of the latter to those of the former is calculated, stored and inputted into a correction device to correct the forecast power distribution obtained by the power distribution calculation device. This enables to forecast the power distribution with excellent responsivity in the reactor site. (Furukawa, Y.)

  19. Alcohol beverage control, privatization and the geographic distribution of alcohol outlets

    Directory of Open Access Journals (Sweden)

    Grubesic Tony H

    2012-11-01

    Full Text Available Abstract Background With Pennsylvania currently considering a move away from an Alcohol Beverage Control state to a privatized alcohol distribution system, this study uses a spatial analytical approach to examine potential impacts of privatization on the number and spatial distribution of alcohol outlets in the city of Philadelphia over a long time horizon. Methods A suite of geospatial data were acquired for Philadelphia, including 1,964 alcohol outlet locations, 569,928 land parcels, and school, church, hospital, park and playground locations. These data were used as inputs for exploratory spatial analysis to estimate the expected number of outlets that would eventually operate in Philadelphia. Constraints included proximity restrictions (based on current ordinances regulating outlet distribution of at least 200 feet between alcohol outlets and at least 300 feet between outlets and schools, churches, hospitals, parks and playgrounds. Results Findings suggest that current state policies on alcohol outlet distributions in Philadelphia are loosely enforced, with many areas exhibiting extremely high spatial densities of outlets that violate existing proximity restrictions. The spatial model indicates that an additional 1,115 outlets could open in Philadelphia if privatization was to occur and current proximity ordinances were maintained. Conclusions The study reveals that spatial analytical approaches can function as an excellent tool for contingency-based “what-if” analysis, providing an objective snapshot of potential policy outcomes prior to implementation. In this case, the likely outcome is a tremendous increase in alcohol outlets in Philadelphia, with concomitant negative health, crime and quality of life outcomes that accompany such an increase.

  20. Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

    Science.gov (United States)

    Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

  1. Hydrologic Outlets of the Greenland Ice Sheet, Version 1

    Data.gov (United States)

    National Aeronautics and Space Administration — The Hydrologic Outlets of the Greenland Ice Sheet data set contains GIS point shapefiles that include 891 observed and potential hydrologic outlets of the Greenland...

  2. nijotech vol. 6 no. 1 september 1982 iloeje 25 effects of parallel

    African Journals Online (AJOL)

    Dr Obe

    upper plenum into the channels and lower plenum of Boiling Water Nuclear Power. Reactors during ... pressures varied from near atmospheric to a little over 1.7 bar. Test section ... The steam generator was an Electro-Magic (Model. 3100) unit ...

  3. The pebble-bed high-temperature reactor as a source of nuclear process heat. Vol. 3

    International Nuclear Information System (INIS)

    Kugeler, K.; Schulten, R.; Kugeler, M.; Niessen, H.F.; Roeth-Kamat, M.; Hohn, H.; Woike, O.; Germer, J.H.

    1974-08-01

    The characteristic questions concerning a process heat reactor with high helium outlet temperatures are dealt with in this volume like e.g. fuel element design, corrosion, and fission product release. Furthermore, some possibilities of the technical realization of the hot-gas ducting and intermediate heat exchangers are described. Important parameters for the design of the reactor such as core power density, helium inlet and outlet temperatures, helium pressure and fuel cycle burn-up and conversion and the effect of these on the primary circuit are investigated. The important question regarding which reactor vessel is to be chosen for nuclear process heat plants is discussed with the aid of the integrated and non-integrated concepts using prestressed concrete, cast iron and cast steel. Thereafter, considerations on the safety of the nuclear plant are given. Finally, mention is made of the availability of the nuclear plant and of the status of development of the HTR technology. (orig.) [de

  4. Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY1999-2001

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    The HTTR (High Temperature Engineering Test Reactor) with the thermal power of 30 MW and the reactor outlet coolant temperature of 850/950 degC is the first high temperature gas-cooled reactor (HTGR) in Japan, which uses coated fuel particle, graphite for core components, and helium gas for primary coolant. The HTTR, which locates at the south-west area of 50,000 m{sup 2} in the Oarai Research Establishment, had been constructed since 1991 before accomplishing the first criticality on November 10, 1998. Rise to power tests of the HTTR started in September, 1999 and the rated thermal power of 30 MW and the reactor outlet coolant temperature of 850 degC was attained in December 2001. JAERI received the certificate of pre-operation test, that is, the commissioning license for the HTTR in March 2002. This report summarizes operation, tests, maintenance, radiation control, and construction of components and facilities for the HTTR as well as R and Ds on HTGRs from FY1999 to 2001. (author)

  5. AGU hydrology publication outlets

    Science.gov (United States)

    Freeze, R. Allan

    In recent months I have been approached on several occasions by members of the hydrology community who asked me which of the various AGU journals and publishing outlets would be most suitable for a particular paper or article that they have prepared.Water Resources Research (WRR) is the primary AGU outlet for research papers in hydrology. It is an interdisciplinary journal that integrates research in the social and natural sciences of water. The editors of WRR invite original contributions in the physical, chemical and biological sciences and also in the social and policy sciences, including economics, systems analysis, sociology, and law. The editor for the physical sciences side of the journal is Donald R. Nielson, LAWR Veihmeyer Hall, University of California Davis, Davis, CA 95616. The editor for the policy sciences side of the journal is Ronald G. Cummings, Department of Economics, University of New Mexico, Albuquerque, NM 87131

  6. The 3D thermal-hydraulic numerical simulation for the fuel zone outlet of China experimental fast reactor

    International Nuclear Information System (INIS)

    Xue Xiuli; Yang Hongyi; Yang Fuchang

    2008-01-01

    Detailed 3D thermal-hydraulic numerical analyses to the fuel zone outlet are actualized with the STAR-CD CFD code. The performance of sodium mixing is studied and detailed velocity and temperature distribution are obtained in this region which will offer foundations and references to study the rationality of temperature monitoring-spot arrangement and to assess the effect of temperature fluctuations to control rod guide tubes in this region, and so on. (authors)

  7. A Small Modular Reactor Design for Multiple Energy Applications: HTR50S

    Energy Technology Data Exchange (ETDEWEB)

    Yan, X.; Tachibana, Y.; Ohashi, H.; Sato, H.; Tazawa, Y.; Kunitomi, K. [Japan Atomic Energy Agency, Ibaraki (Japan)

    2013-06-15

    HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's 950 .deg. C, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to 750 .deg. C for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to 900 .deg. C for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.

  8. Acute Alcohol Consumption, Alcohol Outlets, and Gun Suicide

    Science.gov (United States)

    Branas, Charles C.; Richmond, Therese S.; Ten Have, Thomas R.; Wiebe, Douglas J.

    2014-01-01

    A case–control study of 149 intentionally self-inflicted gun injury cases (including completed gun suicides) and 302 population-based controls was conducted from 2003 to 2006 in a major US city. Two focal independent variables, acute alcohol consumption and alcohol outlet availability, were measured. Conditional logistic regression was adjusted for confounding variables. Gun suicide risk to individuals in areas of high alcohol outlet availability was less than the gun suicide risk they incurred from acute alcohol consumption, especially to excess. This corroborates prior work but also uncovers new information about the relationships between acute alcohol consumption, alcohol outlets, and gun suicide. Study limitations and implications are discussed. PMID:21929327

  9. Gas-cooled reactor for space power systems

    International Nuclear Information System (INIS)

    Walter, C.E.; Pearson, J.S.

    1987-05-01

    Reactor characteristics based on extensive development work on the 500-MWt reactor for the Pluto nuclear ramjet are described for space power systems useful in the range of 2 to 20 MWe for operating times of 1 y. The modest pressure drop through the prismatic ceramic core is supported at the outlet end by a ceramic dome which also serves as a neutron reflector. Three core materials are considered which are useful at temperatures up to about 2000 K. Most of the calculations are based on a beryllium oxide with uranium dioxide core. Reactor control is accomplished by use of a burnable poison, a variable-leakage reflector, and internal control rods. Reactivity swings of 20% are obtained with a dozen internal boron-10 rods for the size cores studied. Criticality calculations were performed using the ALICE Monte Carlo code. The inherent high-temperature capability of the reactor design removes the reactor as a limiting condition on system performance. The low fuel inventories required, particularly for beryllium oxide reactors, make space power systems based on gas-cooled near-thermal reactors a lesser safeguard risk than those based on fast reactors

  10. Predictions of the Bypass Flows in the HTR-PM Reactor Core

    International Nuclear Information System (INIS)

    Sun Jun; Chen Zhipeng; Zheng Yanhua; Shi Lei; Li Fu

    2014-01-01

    In the HTR-PM reactor core, the basic structure materials are large amount of graphite reflectors and carbon bricks. Small gaps among those graphite and carbon bricks are widespread in the reactor core so that the cold helium flow may be bypassed and not completely heated. The bypass flows in relative lower temperature would change the flow and temperature distributions in the reactor core, therefore, the accurate prediction of bypass flows need to be carried out carefully to evaluate the influence to the reactor safety. Based on the characteristics of the bypass flow problem, hybrid method of the flow network and the CFD tools was employed to represent the connections and calculate flow distributions of all the main flow and bypass flow paths. In this paper, the hybrid method was described and applied to specific bypass flow problem in the HTR-PM. Various bypass flow paths in the HTR-PM were reviewed, figured out, and modeled by the flow network and the CFD methods, including the axial vertical gaps in the side reflectors, control rod channels, absorber sphere channels and radial gap flow through keys around the hot helium plenum. The bypass flow distributions and its flow rate ratio to the total flow rate in the primary loop were also calculated, discussed and evaluated. (author)

  11. Design and analysis on super-critical water cooled power reactors

    International Nuclear Information System (INIS)

    Ishiwatari, Yuki

    2005-01-01

    The Super-Critical Water Cooled Power Reactors (SCPR) is cooled by 25 MPa supercritical water of 280degC at reactor inlet and greater than 500degC at reactor outlet and directly connected with turbine/generators with high energy conversion efficiency. This corresponds to the deletion of recirculation system and steam-water separation system of BWR type reactors or of pressurizer and steam generator of PWR type reactors. In addition to the design study of the university of Tokyo, technology development of the SCPR for practical use has started under the collaboration of industry and academia since 2000. Mockup single tube and bundle tests for heat transfer/fluid flow characteristics of the design have been conducted with 3D heat transfer analysis. Materials compatible with coolant conditions for fuel cans and reactor internals are also assessed. Overall evaluation of the reactor concept is under way. (T. Tanaka)

  12. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  13. Thermal-hydraulic instabilities in pressure tube graphite-moderated boiling water reactors

    International Nuclear Information System (INIS)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling charmers in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement

  14. High-temperature reactor in modular construction

    International Nuclear Information System (INIS)

    Mueller, F.U.; Reutler, H.; Ullrich, M.

    1981-01-01

    Together with other reactors of the same type a gas-cooled, small-sized high-temperature reactor is to be assembled into a plant with modular design. The reactor vessel can be withdrawn as a whole after shutdown, removal of the fuel element charge, disassembly of the control rods, and opening of the closure of the safety containment. All apertures for the inlet and outlet of the cooling gas are located in the ground plate of the reactor. The lower part of the reactor cavern serves as inlet space for the cool gas, while the heated gas is let in through a line of a heat sink, e.g. a heat exchanger. The ground plate is connected with the hot gas line or with an inserted hot gas collecting room by means of a simple plug connection which is released automatically when the reactor vessel is withdrawn. The cooling gas, which is put into circulation by a blower and led through special conducting systems, is also used for cooling the outer metal jacket of the hot gas line. A second design is described according to which the reactor and heat exchanger are superposed in a safety containment, such as applied for pressurized water-cooled nuclear reactors. (orig.) [de

  15. The lithium-lithium hydride process for the production of hydrogen: comparison of two concepts for 950 and 1300 deg C HTR helium outlet temperature

    International Nuclear Information System (INIS)

    Oertel, M.; Weirich, W.; Kuegler, B.; Luecke, L.; Pietsch, M.; Winkelmann, U.

    1987-01-01

    The lithium-lithium hydride process serves to generate hydrogen from water efficiently, using the high temperature heat of a nuclear reactor. Thermodynamic analyses show that hydrogen can be produced with an overall thermal efficiency of 48% at conventional HTR outlet temperatures of 950 0 C. Assuming helium heat of 1300 0 C, 56% overall thermal efficiency can be achieved. (author)

  16. MELCOR Severe Accident Analysis on the SMART Reactor

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Jin, Young Ho; Kim, Young In; Kim, Keung Koo; Wang, Ziao; Revankar, Shripad

    2014-01-01

    A severe accident is analyzed for Korea SMR reactor, SMART. Core melt down sequences are analyzed for SMART reactor core using MELCOR version 1.8.5. MELCOR is developed by Sandia National Laboratory for US NRC for the simulation of severe accidents in nuclear power plants. Two cases are simulated here and compared between them; one is the case for core having 3 concentric rings and the other is the case for core having 5 concentric rings. One inch break LOCA scenario is simulated and compared between these two core models. Time sequences for the thermal hydraulic behaviors of RPV and thermal heatup behaviors of reactor core are explained in graphically. Thermal hydraulic behavior such as the change of pressure, level, mass, and temperature of RPV is explained. Thermal heatup behavior of reactor core such as oxidation of cladding, hydrogen generation, core slumping down to lower plenum, and finally creep rupture of PRV lower head is explained. Engineered safety features such as safety injection systems (SIS), and Passive residual heat removal systems (PHRS), etc. are assumed to be not working. One inch break of severe accident is simulated on Korean SMR (SMART) Integral PWR with MELCOR code version 1.8.5. Core melt progression and lower head failure time is very slow compared to other commercial reactors. Simulation on 3 and 5 radial rings core models gives very similar pattern in core cell failure timings. Other various accident scenarios (for example, SBO in Fukushima) will be tried further. Containment behaviors and source term behaviors in severe accident conditions will be analyzed in future

  17. Thermal baffle for fast-breeder reactor

    International Nuclear Information System (INIS)

    Rylatt, J.A.

    1977-01-01

    A liquid-metal-cooled fast-breeder reactor includes a bridge structure for separating hot outlet coolant from relatively cool inlet coolant consisting of an annular stainless steel baffle plate extending between the core barrel surrounding the core and the thermal liner associated with the reactor vessel and resting on ledges thereon, there being inner and outer circumferential webs on the lower surface of the baffle plate and radial webs extending between the circumferential webs, a stainless steel insulating plate completely covering the upper surface of the baffle plate and flex seals between the baffle plate and the ledges on which the baffle plate rests to prevent coolant from washing through the gaps therebetween. The baffle plate is keyed to the core barrel for movement therewith and floating with respect to the thermal liner and reactor vessel. 3 claims, 2 figures

  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. Improvement of seawater booster pump outlet check valve

    International Nuclear Information System (INIS)

    Li Xuning; Du Yansong; Huang Huimin

    2010-01-01

    Conventional island seawater booster pump set of QNPC 310 MWe unit are very important in the whole circulating cooling system, and the integrate function of seawater booster pump outlet check valve is the foundation of steady operation of the seawater booster pump set. The article mainly introduce that through the analyses to the reason to the problem that the seawater booster pump outlet check valve of QNPC 310 MWe unit appeared in past years by our team, and considering the influence of operation condition and circumstance, the team improve the seawater booster pump outlet check valve from swing check valve to shuttle check valve which operate more appropriately in the system. By the test of continuous practice, we make further modification to the inner structure of shuttle check valve contrapuntally, and therefore we solve the problem in seawater booster pump outlet check valve fundamentally which has troubled the security of system operation in past years, so we realize the aim of technical improvement and ensure that the system operate in safety and stability. (authors)

  20. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gougar, Hans D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  1. Baseline Concept Description of a Small Modular High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hans Gougar

    2014-05-01

    The objective of this report is to provide a description of generic small modular high temperature reactors (herein denoted as an smHTR), summarize their distinguishing attributes, and lay out the research and development (R&D) required for commercialization. The generic concepts rely heavily on the modular high temperature gas-cooled reactor designs developed in the 1980s which were never built but for which pre-licensing or certification activities were conducted. The concept matured more recently under the Next Generation Nuclear Plant (NGNP) project, specifically in the areas of fuel and material qualification, methods development, and licensing. As all vendor-specific designs proposed under NGNP were all both ‘small’ or medium-sized and ‘modular’ by International Atomic Energy Agency (IAEA) and Department of Energy (DOE) standards, the technical attributes, challenges, and R&D needs identified, addressed, and documented under NGNP are valid and appropriate in the context of Small Modular Reactor (SMR) applications. Although the term High Temperature Reactor (HTR) is commonly used to denote graphite-moderated, thermal spectrum reactors with coolant temperatures in excess of 650oC at the core outlet, in this report the historical term High Temperature Gas-Cooled Reactor (HTGR) will be used to distinguish the gas-cooled technology described herein from its liquid salt-cooled cousin. Moreover, in this report it is to be understood that the outlet temperature of the helium in an HTGR has an upper limit of 950 degrees C which corresponds to the temperature to which certain alloys are currently being qualified under DOE’s ARC program. Although similar to the HTGR in just about every respect, the Very High Temperature Reactor (VHTR) may have an outlet temperature in excess of 950 degrees C and is therefore farther from commercialization because of the challenges posed to materials exposed to these temperatures. The VHTR is the focus of R&D under the

  2. 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)

  3. Holding device for gas-cooled reactor fuel elements

    International Nuclear Information System (INIS)

    Hensolt, T.

    1980-01-01

    The sheathed fuel elements of the GCFR are inserted with their pedestal in a grid plate arranged below the reactor core and are clamped there. The clamping force as well as the force required for hydraulic holding-down against the flow pressure of the coolant are applied through the differential pressure between inlet and outlet of the coolant. (DG) [de

  4. Moderator heat recovery of CANDU reactors

    International Nuclear Information System (INIS)

    Fath, H.E.S.; Ahmed, S.T.

    1986-01-01

    A moderator heat recovery scheme is proposed for CANDU reactors. The proposed circuit utilizes all the moderator heat to the first stages of the plant feedwater heating system. CANDU-600 reactors are considered with moderator heat load varying from 120 to 160 MWsub(th), and moderator outlet temperature (from calandria) varying from 80 to 100 0 C. The steam saved from the turbine extraction system was found to produce an additional electric power ranging from 5 to 11 MW. This additional power represents a 0.7-1.7% increase in the plant electric output power and a 0.2-0.7% increase in the plant thermal efficiency. The outstanding features and advantages of the proposed scheme are presented. (author)

  5. CFD simulation analysis and validation for CPR1000 pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Mingqian; Ran Xiaobing; Liu Yanwu; Yu Xiaolei; Zhu Mingli

    2013-01-01

    Background: With the rapid growth in the non-nuclear area for industrial use of Computational fluid dynamics (CFD) which has been accompanied by dramatically enhanced computing power, the application of CFD methods to problems relating to Nuclear Reactor Safety (NRS) is rapidly accelerating. Existing research data have shown that CFD methods could predict accurately the pressure field and the flow repartition in reactor lower plenum. But simulations for the full domain of the reactor have not been reported so far. Purpose: The aim is to determine the capabilities of the codes to model accurately the physical phenomena which occur in the full reactor vessel. Methods: The flow field of the CPR1000 reactor which is associated with a typical pressurized water reactor (PWR) is simulated by using ANSYS CFX. The pressure loss in reactor pressure vessel, the hydraulic loads of guide tubes and support columns, and the bypass flow of head dome were obtained by calculations for the full domain of the reactor. The results were validated by comparing with the determined reference value of the operating nuclear plant (LingAo nuclear plant), and the transient simulation was conducted in order to better understand the flow in reactor pressure vessel. Results: It was shown that the predicted pressure loss with CFD code was slightly different with the determined value (10% relative deviation for the total pressure loss), the hydraulic loads were less than the determined value with maximum relative deviation 50%, and bypass flow of head dome was approximately the same with determined value. Conclusion: This analysis practice predicts accurately the physical phenomena which occur in the full reactor vessel, and can be taken as a guidance for the nuclear plant design development and improve our understanding of reactor flow phenomena. (authors)

  6. A ceramic breeder in a poloidal tube blanket for a tokamak reactor

    International Nuclear Information System (INIS)

    Amici, A.; Anzidei, L.; Gallina, M.; Rado, V.; Simbolotti, G.; Violante, V.; Zampaglione, V.; Petrizzi, L.

    1989-01-01

    A conceptual study of a helium-cooled solid breeder blanket for a tokamak reactor is presented. Tritium breeding capability together with system reliability are taken as the main design criteria. The blanket consists of tubular poloidal modules made of a central bundle of ceramic rods (γLiAlO 2 ) with a coaxial distribution of the inlet/outlet coolant flow (He) surrounded by a multiplier material (Be) in the form of bored bricks. The Be to γLiAlO 2 volume ratio is 4/1. The He inlet and outlet branches are cooling Be and γLiAlO 2 , respectively. A purge He flow running through small central holes of the ceramic rods is derived from the main flow. Under the typical conditions of a tokamak reactor (neutron wall load=2 MW/m 2 ), a full coverage tritium breeding ratio of 1.47 is achieved for the following design and operating parameters: outlet He temperature=570 0 C; inlet He temperature=250 0 ; total extracted power=2700 MW; He pumping power percentage=2%; minimum/maximum γLiAlO 2 temperature=400/900 0 C; maximum structural temperature=475 0 C; and maximum Be temperature=525 0 C. (orig.)

  7. Development of high temperature gas cooled reactor in China

    Energy Technology Data Exchange (ETDEWEB)

    Guo, Wentao [Paul Scherrer Institute, Villigen (Switzerland). Dept. of Nuclear Energy and Safety; Schorer, Michael [Swiss Nuclear Forum, Olten (Switzerland)

    2018-02-15

    High temperature gas cooled reactor (HTGR) is one of the six Generation IV reactor types put forward by Generation IV International Forum (GIF) in 2002. This type of reactor has high outlet temperature. It uses Helium as coolant and graphite as moderator. Pebble fuel and ceramic reactor core are adopted. Inherit safety, good economy, high generating efficiency are the advantages of HTGR. According to the comprehensive evaluation from the international nuclear community, HTGR has already been given the priority to the research and development for commercial use. A demonstration project of the High Temperature Reactor-Pebble-�bed Modules (HTR-PM) in Shidao Bay nuclear power plant in China is under construction. In this paper, the development history of HTGR in China and the current situation of HTR-PM will be introduced. The experiences from China may be taken as a reference by the international nuclear community.

  8. A longitudinal analysis of alcohol outlet density and domestic violence.

    Science.gov (United States)

    Livingston, Michael

    2011-05-01

    A small number of studies have identified a positive relationship between alcohol outlet density and domestic violence. These studies have all been based on cross-sectional data and have been limited to the assessment of ecological correlations between outlet density and domestic violence rates. This study provides the first longitudinal examination of this relationship. Cross-sectional time-series using aggregated data from small areas. The relationships between alcohol outlet density and domestic violence were assessed over time using a fixed-effects model. Controls for the spatial autocorrelation of the data were included in the model. The study uses data for 186 postcodes from within the metropolitan area of Melbourne, Australia for the years 1996 to 2005. Alcohol outlet density measures for three different types of outlets (hotel/pub, packaged liquor, on-premise) were derived from liquor licensing records and domestic violence rates were calculated from police-recorded crime data, based on the victim's postcode. Alcohol outlet density was associated significantly with rates of domestic violence, over time. All three licence categories were positively associated with domestic violence rates, with small effects for general (pub) and on-premise licences and a large effect for packaged liquor licences. In Melbourne, the density of liquor licences is positively associated with rates of domestic violence over time. The effects were particularly large for packaged liquor outlets, suggesting a need for licensing policies that pay more attention to o off-premise alcohol availability. © 2011 The Authors, Addiction © 2011 Society for the Study of Addiction.

  9. Thermal and stability considerations for a supercritical water-cooled fast reactor during power-raising phase of plant startup

    International Nuclear Information System (INIS)

    Cai, Jiejin; Ishiwatari, Yuki; Oka, Yoshiaki; Ikejiri, Satoshi

    2009-01-01

    This paper describes thermal analyses and linear stability analyses of the Supercritical Water-cooled Fast Reactor with 'two-path' flow scheme during the power-raising phase of plant startup. For thermal consideration, the same criterion of the maximum cladding surface temperature (MCST) as applied to the normal operating condition is used. For thermal-hydraulic stability consideration, the decay ratio of 0.5 is applied, which is taken from BWRs. Firstly, we calculated the flow rate distribution among the parallel flow paths from the reactor vessel inlet nozzles to the mixing plenum below the core using a system analysis code. The parallel flow paths consist of the seed fuel assemblies cooled by downward flow, the blanket fuel assemblies cooled by downward flow and the downcomer. Then, the MCSTs are estimated for various reactor powers and feedwater flow rates with system analyses. The decay ratios are estimated with linear stability analyses. The available range of the reactor power and feedwater flow rate to satisfy the thermal and stability criteria is obtained. (author)

  10. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Gatley, J.A.

    1979-01-01

    Breeder fuel sub-assemblies with electromagnetic brakes and fluidic valves for liquid metal cooled fast breeder reactors are described. The electromagnetic brakes are of relatively small proportions and the valves are of the controlled vortex type. The outlet coolant temperature of at least some of the breeder sub-assemblies are maintained by these means substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (UK)

  11. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Thatcher, G.; Mitchell, A.J.

    1981-01-01

    Fuel sub-assemblies for liquid metal-cooled fast breeder reactors are described which each incorporate a fluid flow control valve for regulating the rate of flow through the sub-assembly. These small electro-magnetic valves seek to maintain the outlet coolant temperature of at least some of the breeder sub-assemblies substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (U.K.)

  12. The experimental study on the mass transfer model of boron injection for natural circular heating reactor

    International Nuclear Information System (INIS)

    Zha Meisheng; Nie Mengchen; Zhou Huizhong; Wang Liqun; Guo Weiping; Liu ZHiyong

    1989-09-01

    A pulse injection stimulus-response technique to study the boron mixing and transport performance after boron-loaded liquid was injected into the reactor core is described. The experiment was carried out in a simulation device. The simulation medium was used. The experimental results show that the lower plenum where the injection point located can be simplified to one scale inertial unit and the movement of boron mixture was only transported after it had entered into the fuel elements. The definition of boron initiative mixing fraction η is also given. By using relating data a dimensionless equation is obtained

  13. Gastric Outlet Obstruction from Duodenal Lipoma in an Adult ...

    African Journals Online (AJOL)

    Gastric Outlet Obstruction from Duodenal Lipoma in an Adult. ... Nigerian Journal of Surgery ... Although, peptic ulcer disease remains the most common benign cause of gastric outlet obstruction (GOO), duodenal lipomas remain a rare, but possible cause of GOO and could pose a diagnostic challenge, especially in ...

  14. Facilities of fuel transfer for nuclear reactors

    International Nuclear Information System (INIS)

    Wade, E.E.

    1977-01-01

    This invention relates to sodium cooled fast breeder reactors. It particularly concerns facilities for the transfer of fuel assemblies between the reactor core and a fuel transfer area. The installation is simple in construction and enables a relatively small vessel to be used. In greater detail, the invention includes a vessel with a head, fuel assemblies housed in this vessel, and an inlet and outlet for the coolant covering these fuel assemblies. The reactor has a fuel transfer area in communication with this vessel and gear inside the vessel for the transfer of these fuel assemblies. These facilities are borne by the vessel head and serve to transfer the fuel assemblies from the vessel to the transfer area; whilst leaving the fuel assemblies completely immersed in a continuous mass of coolant. A passageway is provided between the vessel and this transfer area for the fuel assemblies. Facilities are provided for closing off this passageway so that the inside of the reactor vessel may be isolated as desired from this fuel transfer area whilst the reactor is operating [fr

  15. The HTR-10 test reactor project and potential use of HTGR for non-electric application in China

    International Nuclear Information System (INIS)

    Sun Yuliang; Zhong Daxin; Xu Yuanhui; Wu Zhongxin

    1997-01-01

    Coal is the dominant source of energy in China. This use of coal results in two significant problems for China; it is a major burden on the train, road and waterway transportation infrastructures and it is a significant source of environmental pollution. In order to ease the problems caused by the burning of coal and to help reduce the energy supply shortage in China, national policy has directed the development of nuclear power. This includes the erection of nuclear power plants with water cooled reactors and the development of advanced nuclear reactor types, specifically, the high temperature gas cooled reactor (HTGR). The HTGR was chosen for its favorable safety features and its ability to provide high reactor outlet coolant temperatures for efficient power generation and high quality process heat for industrial applications. As the initial modular HTGR development activity within the Chinese High Technology Programme, a 10MW helium cooled test reactor is currently under construction on the site of the Institute of Nuclear Energy Technology northwest of Beijing. This plant features a pebble-bed helium cooled reactor with initial criticality anticipated in 1999. There will be two phases of high temperature heat utilization from the HTR-10. The first phase will utilize a reactor outlet temperature of 700 deg. C with a steam generator providing steam for a steam turbine cycle which works on an electrical/heat co-generation basis. The second phase is planned for a core outlet temperature of 900 deg. C to investigate a steam cycle/gas turbine combined cycle system with the gas turbine and the steam cycle being independently parallel in the secondary side of the plant. This paper provides a review of the technical design, licensing, safety and construction schedule for the HTR-10. It also addresses the potential uses of the HTGR for non-electric applications in China including process steam for the petrochemical industry, heavy oil recovery, coal conversion and

  16. Nuclear reactor fuel rod

    International Nuclear Information System (INIS)

    Busch, H.; Mindnich, F.R.

    1973-01-01

    The fuel rod consists of a can with at least one end cap and a plenum spring between this cap and the fuel. To prevent the hazard that a eutectic mixture is formed during welding of the end cap, a thermal insulation is added between the end cap and plenum spring. It consists of a comical extension of the end cap with a terminal disc against which the spring is supported. The end cap, the extension, and the disc may be formed by one or several pieces. If the disc is separated from the other parts it may be manufactured from chrome steel or VA steel. (DG) [de

  17. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    1975-01-01

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

  18. Seals in nuclear reactors

    International Nuclear Information System (INIS)

    1979-01-01

    The seals described are for use in a nuclear reactor where there are fuel assemblies in a vessel, an inlet and an outlet for circulating a coolant in heat transfer relationship with the fuel assemblies and a closure head on the vessel in a tight fluid relationship. The closure head comprises rotatable plugs which have mechanical seals disposed in the annulus around each plug while allowing free rotation of the plug when the seal is not actuated. The seal is usually an elastomer or copper. A means of actuating the seal is attached for drawing it vertically into the annulus for sealing. When the reactor coolant is liquid sodium, contact with oxygen must be avoided and argon cover gas fills the space between the bottom of the closure head and the coolant liquid level and the annuli in the closure head. (U.K.)

  19. 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

  20. Measurement on the effect of sound wave in upper plenum of boiling water reactor

    International Nuclear Information System (INIS)

    Kumagai, Kosuke; Someya, Satoshi; Okamoto, Koji

    2009-01-01

    In recent years, the power uprate of Boiling Water Reactors have been conducted at several existing power plants as a way to improve plant economy. In one of the power uprated plants (117.8% uprates) in the United States, the steam dryer breakages due to fatigue fracture occurred. It is conceivable that the increased steam flow passing through the branches caused a self-induced vibration with the propagation of sound wave into the steam-dome. The resonance among the structure, flow and the pressure fluctuation resulted in the breakages. To understand the basic mechanism of the resonance, previous researches were done by a point measurement of the pressure and by a phase averaged measurement of the flow, while it was difficult to detect the interaction among them by the conventional method. In this study, Dynamic Particle Image Velocimetry (PIV) System was applied to investigate the effect of sound on natural convection and forced convection. Especially, when the phases of acoustic sources were different, various acoustic wave effects were checked. (author)

  1. Neighbourhood fast food outlets and obesity in children and adults: the CLAN Study.

    Science.gov (United States)

    Crawford, David A; Timperio, Anna F; Salmon, Jo A; Baur, Louise; Giles-Corti, Billie; Roberts, Rebecca J; Jackson, Michelle L; Andrianopoulos, Nick; Ball, Kylie

    2008-01-01

    We examined associations between density of and proximity to fast food outlets and body weight in a sample of children (137 aged 8-9 years and 243 aged 13-15 years) and their parents (322 fathers and 362 mothers). Children's measured and parents' self-reported heights and weights were used to calculate body mass index (BMI). Locations of major fast food outlets were geocoded. Bivariate linear regression analyses examined associations between the presence of any fast food outlet within a 2 km buffer around participants' homes, fast food outlet density within the 2 km buffer, and distance to the nearest outlet and BMI. Each independent variable was also entered into separate bivariate logistic regression analyses to predict the odds of being overweight or obese. Among older children, those with at least one outlet within 2 km had lower BMI z-scores. The further that fathers lived from an outlet, the higher their BMI. Among 13-15-year-old girls and their fathers, the likelihood of overweight/obesity was reduced by 80% and 50%, respectively, if they had at least one fast food outlet within 2 km of their home. Among older girls, the likelihood of being overweight/obese was reduced by 14% with each additional outlet within 2 km. Fathers' odds of being overweight/obese increased by 13% for each additional kilometre to the nearest outlet. While consumption of fast food has been shown to be associated with obesity, this study provides little support for the concept that exposure to fast food outlets in the local neighbourhood increases risk of obesity.

  2. Reactor shut-down device

    International Nuclear Information System (INIS)

    Otsuka, Fumio; Horikawa, Yuji.

    1990-01-01

    The present invention concerns an externally disposed reactor shut-down device for an FBR type reactor using liquid sodium as coolants. An introducing pipe having an outlet port disposed at an upper portion thereof is disposed at a lower end of an upper guide tube. An extension tube, an L-shaped measuring wire support and a measuring wire are disposed at the inside of the guide tube. With such a constitution, low temperature coolants flown out from the lower guide tube of a control rod and a great amount of high temperature coolants flown out from the lower guide tube of a fuel assembly are introduced smoothly to the introducing tube having the measuring wire support disposed therein. Accordingly, the high temperature coolants can be prevented from flowing out to the outside of the introducing tube and coolants after mixing can be flown and hit against a curie point electromagnet efficiently. This can make the response to abnormal temperature rise of coolants satisfactory and can provide reliable reactor scram. (I.N.)

  3. Pump/heat exchanger assembly for pool-type reactor

    International Nuclear Information System (INIS)

    Nathenson, R.D.; Slepian, R.M.

    1987-01-01

    A heat exchanger and pump assembly comprising a heat exchanger including a housing for defining an annularly shaped cavity and supporting therein a plurality of heat transfer tubes. A pump is disposed beneath the heat exchanger and is comprised of a plurality of flow couplers disposed in a circular array. Each flow coupler is comprised of a pump duct for receiving a first electrically conductive fluid, i.e. the primary liquid metal, from a pool thereof, and a generator duct for receiving a second electrically conductive fluid, i.e. the intermediate liquid metal. The primary liquid metal is introduced from the reactor pool into the top, inlet ends of the tubes, flowing downward therethrough to be discharged from the tubes' bottom ends directly into the reactor pool. The primary liquid metal is variously introduced into the pump ducts directly from the reactor pool, either from the bottom or top end of the flow coupler. The intermediate fluid introduced into the generator ducts via the inlet duct and inlet plenum and after leaving the generator ducts passes through the annular cavity of the exchanger to cool the primary liquid in the tubes. The annular magnetic field of the pump is produced by a circular array of electromagnets having hollow windings cooled by a flow of the intermediate metal. (author)

  4. Evaluation of alcohol outlet density and its relation with violence

    Directory of Open Access Journals (Sweden)

    Laranjeira Ronaldo

    2002-01-01

    Full Text Available OBJECTIVES: The current study set out to investigate alcohol availability in a densely populated, residential area of suburban São Paulo associated with high levels of social deprivation and violence. Gun-related deaths and a heavy concentration of alcohol outlets are notable features of the area surveyed. Given the strong evidence for a link between alcohol availability and a number of alcohol-related problems, including violent crime, measures designed to reduce accessibility have become a favored choice for alcohol prevention programs in recent years. METHODS: The interviewers were 24 residents of the area who were trained for the study. It was selected an area of nineteen streets, covering a total distance of 3.7 km. A profile of each alcohol outlet available on the area was recorded. RESULTS: One hundred and seven alcohol outlets were recorded. The number of other properties in the same area was counted at 1,202. Two measures of outlet density may thus be calculated: the number of outlets per kilometer of roadway (29 outlets/km; and the proportion of all properties that sold alcohol (1 in 12. CONCLUSIONS: The results of this study is compared with others which are mainly from developed countries and shown that the area studied have the highest density of alcohol outlet density ever recorded in the medical literature. The implication of this data related to the violence of the region is discussed. By generating a profile of alcohol sales and selling points, it was hoped to gain a better understanding of alcohol access issues within the sample area. Future alcohol prevention policy would be well served by such knowledge.

  5. Method and device for controlling ECCS

    International Nuclear Information System (INIS)

    Ikeda, Takashi; Kataoka, Yoshiyuki; Murase, Michio.

    1986-01-01

    Purpose: To accomplish reactor cooling without exposing fuel assemblies out of the coolant and also without inducing counter-current flow (CCFL) brake likely to be caused by excess injection of water even in case of malfunction of one system in a loss-of-coolant accident. Method: In a BWR type reactor having more than two independent ECCS, the lower plenum water level is measured and when the lower plenum is full of water, the ECCS are kept in a fully closed state, and reversely when the lower plenum is not full of water, more coolant than the lost quantity of water will be injected into the plenum at a higher pressure than a pressure at which fuel rods just begin to be exposed to the steam phase. The subcool energy of the emergency coolant to be injected is determined by the decay heat of the core and the change rate of a container pressure. However, the quantity of the emergency coolant is controlled such that the subcool energy will always become less than the overheating energy in the core range and the lower plenum range, thus improving safety and enabling the removal of a prior-art ECCS. (Kamimura, M.)

  6. Once-through cycle, supercritical-pressure light water cooled reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Y.; Koshizuka, S. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab

    2001-07-01

    Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

  7. Once-through cycle, supercritical-pressure light water cooled reactor concept

    International Nuclear Information System (INIS)

    Oka, Y.; Koshizuka, S.

    2001-01-01

    Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

  8. Device for supporting a nuclear reactor core

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The core of a light-water reactor which is enclosed in a prestressed concrete pressure vessel and held within a diffuser basket is supported by a device consisting of a cylindrical shell which surrounds the basket and is rigidly fixed to a plurality of frusto-conical skirts having concurrent axes and located substantially at right angles to the axis of the reactor core. The small base of each skirt is rigidly fixed to the shell and the large base is anchored in openings formed in the reactor vessel for the penetration of coolant inlet and outlet pipes. The top portion of the shell is secured to the top portion of the diffuser basket, a flat surface being formed on the shell at the point of connection with each frusto-conical skirt so as to ensure rigid suspension while permitting thermal expansion

  9. Ectopic pancreas causing partial gastric outlet obstruction: a case ...

    African Journals Online (AJOL)

    Ectopic pancreas causing partial gastric outlet obstruction: a case report and review of literature. ... Nigerian Journal of Surgery ... Gastric outlet obstruction resulting from ectopic pancreas in an adult is the first of its kind in our center; we, therefore, present this case to describe the challenges faced with diagnosis, treatment, ...

  10. Ectopic Pancreas Causing Partial Gastric Outlet Obstruction: A Case ...

    African Journals Online (AJOL)

    Ectopic pancreas is a rare cause of gastric outlet obstruction, perhaps rarer still among Africans. Although the entity is known, the diagnostic challenges are enormous, especially in the poor‑resource environment. Gastric outlet obstruction resulting from ectopic pancreas in an adult is the first of its kind in our center;.

  11. Laboratory services series: an electrical outlet and corded equipment inspection program

    International Nuclear Information System (INIS)

    Davis, E.A.

    1976-04-01

    A research and development laboratory has thousands of electrical outlets providing power to laboratories, offices, shops, and service areas. These outlets provide power for a wide variety of portable equipment and tools that are equipped with cord and plug. Electric safety requires a periodic check of outlet grounding capability and continuing inspection and repair of corded equipment. Personnel, equipment, reports, procedures, and schedule requirements are reported

  12. Nuclear reactor instrumentation

    International Nuclear Information System (INIS)

    Duncombe, E.; McGonigal, G.

    1976-01-01

    Reference is made to the instrumentation of liquid metal cooled fast reactors. In order to ensure the safe operation of such reactors it is necessary to constantly monitor the coolant flowing through the fuel assemblies for temperature and rate of flow, requiring a large number of sensors. An improved and simplified arrangement is claimed in which the fuel assemblies feed a fraction of coolant to three instrument units arranged to sense the temperature and rate of flow of samples of coolant. Each instrument unit comprises a sleeve housing a sensing unit and has a number of inlet ducts arranged for receiving coolant from a fuel assembly together with a single outlet. The sensing unit has three thermocouple hot junctions connected in series, the hot junctions and inlet ducts being arranged in pairs. Electromagnetic windings around an inductive core are arranged to sense variation in flow of liquid metal by flux distortion. Fission product sensing means may also be provided. Full constructional details are given. (U.K.)

  13. DEM study of granular discharge rate through a vertical pipe with a bend outlet in small absorber sphere system

    Energy Technology Data Exchange (ETDEWEB)

    Li, Tianjin, E-mail: tjli@tsinghua.edu.cn; Zhang, He; Liu, Malin; Huang, Zhiyong; Bo, Hanliang; Dong, Yujie

    2017-04-01

    Highlights: • The work concerns granular flow in a vertical pipe with a bend. • Discharge rate fluctuation in vertical pipe are mainly from velocity fluctuation. • Steady discharge rate decreases rapidly and saturates with μ{sub s} increasing. • Steady discharge rate W{sub s} still obey the 5/2 power law of pipe internal diameter. • A correlation developed for steady discharge rate for this new geometry. - Abstract: Absorber sphere pneumatic conveying is a special application of pneumatic conveying technique in the pebble bed High Temperature Gas-Cooled Reactor (HTGR or HTR). Granular discharge through a vertical pipe with a bend outlet is one of the control modes to determine solid mass flowrate which is an important parameter for the design of absorber sphere pneumatic conveying. Granular discharge rate through the vertical pipe with a bend outlet in the small absorber sphere system are investigated by discrete element method simulation. The effect of geometry parameters on discharge rate, the discharge rate fluctuation in the vertical pipe, and the effect of friction on steady discharge rate (W{sub s}) are analyzed and discussed. The phenomena of discharge rate fluctuation in the vertical pipe are observed, which are mainly resulted from the evolution of the average downward granular velocity. The steady discharge rate decreases rapidly with sliding friction coefficient increasing from 0.125 to 0.5, and gradually saturates with the friction coefficient further increasing from 0.5 to 1. It is interesting that the linear relation between W{sub s}{sup 2/5} and pipe internal diameter D with zero intercept are found for the vertical pipe discharge with a bend outlet, which is different from the orifice discharge through a hopper or silo with none-zero intercept. A correlation similar to Beverloo’s correlation is developed to predict the steady discharge rate through the vertical pipe with a bend outlet. These results are helpful for the design of sphere

  14. Alcohol outlet density and alcohol consumption in Los Angeles county and southern Louisiana

    Directory of Open Access Journals (Sweden)

    Matthias Schonlau

    2008-11-01

    Full Text Available The objective of this study was to assess the relationship between alcohol availability, as measured by the density of off-premise alcohol outlets, and alcohol consumption in Los Angeles county and southern Louisiana, USA. Consumption information was collected through a telephone survey of 2,881 households in Los Angeles county and pre-Katrina southern Louisiana, nested within 220 census tracts. Respondents’ addresses were geo-coded and both neighbourhood (census tracts and buffers of varying sizes and individual (network distance to the closest alcohol outlet estimates of off-sale alcohol outlet density were computed. Alcohol outlet density was not associated with the percentage of people who were drinkers in either site. Alcohol outlet density was associated with the quantity of consumption among drinkers in Louisiana but not in Los Angeles. Outlet density within a one-mile buffer of the respondent’s home was more strongly associated with alcohol consumption than outlet density in the respondent’s census tract. The conclusion is that the relationship between neighbourhood alcohol outlet density and alcohol consumption is complex and may vary due to differences in neighbourhood design and travel patterns.

  15. A study on the temperature distribution in the hot leg pipe

    International Nuclear Information System (INIS)

    Choe, Yoon-Jae; Baik, Se-Jin; Jang, Ho-Cheol; Lee, Byung-Jin; Im, In-Young; Ro, Tae-Sun

    2003-01-01

    In the hot leg pipes of reactor coolant system of the Korean Standard Nuclear Power Plant (KSNP), a non-uniform distribution in temperature has been observed across the cross-section, which is attributed to the non-uniformity of power distribution in the reactor core usually having a peak in the center region, and to the colder coolant bypass flow through the reactor vessel outlet nozzle clearances. As a result, the arithmetic mean temperature of four Resistance Temperature Detectors (RTDs) installed in each hot leg - two in the upper region and two in the lower region around the pipe wall may not correctly represent the actual coolant bulk temperature. It is also believed that there is a skewness in the velocity profile in the hot leg pipe due to the sudden changes in the flow direction and area from the core to the hot leg pipe, through the reactor vessel outlet plenum. These temperature non-uniformity and velocity skewness affect the measurement of the plant parameter such as the reactor coolant flow rate which is calculated by using the bulk temperature of hot leg pipes. A computational analysis has been performed to simulate the temperature and velocity distributions and to evaluate the uncertainty of temperature correction offset in the hot leg pipe. A commercial CFD code, FLUENT, is used for this analysis. The analysis results are compared with the operational data of KSNP and the scaled-down model test data for System 80. From the comparisons, an uncertainty of correction offset is obtained to measure the bulk temperature of hot leg more accurately, which can be also applied to the operating plants, leading to the reduction of temperature measurement uncertainty. Since the uncertainty of temperature in the hot leg pipe is one of major parameters to calculate the uncertainty of the reactor coolant flow rate, the analysis results can contribute to the improvement of the plant performance and safety by reducing the uncertainty of temperature measurement

  16. Effects of Parallel Channel Interactions, Steam Flow, Liquid Subcool ...

    African Journals Online (AJOL)

    Tests were performed to examine the effects of parallel channel interactions, steam flow, liquid subcool and channel heat addition on the delivery of liquid from the upper plenum into the channels and lower plenum of Boiling Water Nuclear Power Reactors during reflood transients. Early liquid delivery into the channels, ...

  17. Out-of-home food outlets and area deprivation: case study in Glasgow, UK

    Directory of Open Access Journals (Sweden)

    Cummins Steven

    2005-10-01

    Full Text Available Abstract Background There is a popular belief that out-of-home eating outlets, which typically serve energy dense food, may be more commonly found in more deprived areas and that this may contribute to higher rates of obesity and related diseases in such areas. Methods We obtained a list of all 1301 out-of-home eating outlets in Glasgow, UK, in 2003 and mapped these at unit postcode level. We categorised them into quintiles of area deprivation using the 2004 Scottish Index of Multiple Deprivation and computed mean density of types of outlet (restaurants, fast food restaurants, cafes and takeaways, and all types combined, per 1000 population. We also estimated odds ratios for the presence of any outlets in small areas within the quintiles. Results The density of outlets, and the likelihood of having any outlets, was highest in the second most affluent quintile (Q2 and lowest in the second most deprived quintile (Q4. Mean outlets per 1,000 were 4.02 in Q2, 1.20 in Q4 and 2.03 in Q5. With Q2 as the reference, Odds Ratios for having any outlets were 0.52 (CI 0.32–0.84 in Q1, 0.50 (CI 0.31 – 0.80 in Q4 and 0.61 (CI 0.38 – 0.98 in Q5. Outlets were located in the City Centre, West End, and along arterial roads. Conclusion In Glasgow those living in poorer areas are not more likely to be exposed to out-of-home eating outlets in their neighbourhoods. Health improvement policies need to be based on empirical evidence about the location of fast food outlets in specific national and local contexts, rather than on popular 'factoids'.

  18. Safety design of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)

    International Nuclear Information System (INIS)

    Takahashi, Minoru; Uchida, Shoji; Yamada, Yumi; Koyama, Kazuya

    2008-01-01

    In Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb-Bi coolant above the core, and Pb-Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb-Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. (author)

  19. Design and Optimization for the Windowless Target of the China Nuclear Waste Transmutation Reactor

    Directory of Open Access Journals (Sweden)

    Desheng Cheng

    2016-04-01

    Full Text Available A windowless spallation target can provide a neutron source and maintain neutron chain reaction for a subcritical reactor, and is a key component of China's nuclear waste transmutation of coupling accelerator and subcritical reactor. The main issue of the windowless target design is to form a stable and controllable free surface that can ensure that energy spectrum distribution is acquired for the neutron physical design when the high energy proton beam beats the lead–bismuth eutectic in the spallation target area. In this study, morphology and flow characteristics of the free surface of the windowless target were analyzed through the volume of fluid model using computational fluid dynamics simulation, and the results show that the outlet cross section size of the target is the key to form a stable and controllable free surface, as well as the outlet with an arc transition. The optimization parameter of the target design, in which the radius of outlet cross section is 60 ± 1 mm, is verified to form a stable and controllable free surface and to reduce the formation of air bubbles. This work can function as a reference for carrying out engineering design of windowless target and for verification experiments.

  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. Dynamic simulation of the 2 MWt slowpoke heating reactor

    International Nuclear Information System (INIS)

    Tseng, C.M.; Lepp, R.M.

    1982-04-01

    A 2 MWt SLOWPOKE reactor, intended for commercial space heating, is being developed at the Chalk River Nuclear Laboratories. A small-signal dynamic simulation of this reactor, without closed-loop control, was developed. Basic equations were used to describe the physical phenomena in each kf the eight reactor subsystems. These equations were then linearized about the normal operation conditions and rearranged in a dimensionless form for implementation. The overall simulation is non-linear. Slow transient responses (minutes to days) of the simulation to both reactivity and temperature perturbations were measured at full power. In all cases the system reached a new steady state in times varying from 12 h to 250 h. These results illustrate the benefits of the inherent negative reactivity feedback of this reactor concept. The addition of closed-loop control using core outlet temperature as the controlled variable to move a beryllium reflector is also examined

  2. High temperature and high performance light water cooled reactors operating at supercritical pressure, research and development

    International Nuclear Information System (INIS)

    Oka, Y.; Koshizuka, S.; Katsumura, Y.; Yamada, K.; Shiga, S.; Moriya, K.; Yoshida, S.; Takahashi, H.

    2003-01-01

    The concept of supercritical-pressure, once-through coolant cycle nuclear power plant (SCR) was developed at the University of Tokyo. The research and development (R and D) started worldwide. This paper summarized the conceptual design and R and D in Japan. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical fossil fired power plants (FPP) in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil fired power plants will be fully utilized for SCR. The high temperature, supercritical-pressure light water reactor is the logical evolution of LWR. Boiling evolved from circular boilers, water tube boilers and once-through boilers. It is the reactor version of the once-through boiler. The development from LWR to SCR follows the history of boilers. The goal of the R and D should be the capital cost reduction that cannot be achieved by the improvement of LWR. The reactor can be used for hydrogen production either by catalysis and chemical decomposition of low quality hydrocarbons in supercritical water. The reactor is compatible with tight lattice fast core for breeders due to low outlet coolant density, small coolant flow rate and high head coolant pumps

  3. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the

  4. Safety analysis of switching between reductive and oxidative conditions in a reaction coupling reverse flow reactor.

    NARCIS (Netherlands)

    van Sint Annaland, M.; Kuipers, J.A.M.; van Swaaij, Willibrordus Petrus Maria

    2001-01-01

    A new reverse flow reactor is developed where endothermic reactants (propane dehydrogenation) and exothermic reactants (fuel combustion) are fed sequentially to a monolithic catalyst, while periodically alternating the inlet and outlet positions. Upon switching from reductive to oxidative conditions

  5. GARLIC-B. A digital code for real-time calculation of the transient behaviour of nodal and global core and plant parameters of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Ercan, Y.; Hoeld, A.; Lupas, O.

    1982-04-01

    A program description of the code GARLIC-B is given. The code is based on a nonlinear transient model for BWR nuclear power plants which consist of a 3D-core, a top plenum, steam removal and feed water systems and a downcomer with main coolant recirculation pumps. The core is subdivided into a number of superboxes and flow channels with different coolant mass flow rates. Subcooled boiling within these channels has an important reactivity feed back effect and has to be taken also into account. The code computes the local and global core and plant transient situation as dependent on both the inherent core dynamics and external control actions, i.e., disturbances such as motions of control rod banks, changes of mass flow rates of coolant, feed water and steam outlet. The case of a pressure-controlled reactor operation is also considered. (orig./GL) [de

  6. Application of conjugate gradient method to Commix-1B three-dimensional momentum equation

    International Nuclear Information System (INIS)

    King, J.B.; Domanus, H.

    1987-01-01

    Conjugate gradient method which is a special case of the variational method was implemented in the momentum section of the COMMIX-1B thermal hydraulics code. The comparisons between this method and the conventional iterative method of Successive Over Relation (S.O.R.) were made. Using COMMIX-1B, three steady state problems were analyzed. These problems were flow distribution in a scaled model of the Clinch River Fast Breeder Reactor outlet plenum, flow of coolant in the cold leg and downcomer of a PWR and isothermal air flow through a partially blocked pipe. It was found that if the conjugate gradient method is used, the execution time required to solve the resulting COMMIX-1B system of equations can be reduced by a factor of about 2 for the first two problems. For the isothermal air flow problem, the conjugate gradient method did not improve the execution time

  7. Supercritical-pressure, once-through cycle light water cooled reactor concept

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Koshizuka, Seiichi

    2001-01-01

    The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density. (author)

  8. Water inlet and steam outlet pipes fitted one inside the other for nuclear reactors

    International Nuclear Information System (INIS)

    Mc Donald, B.N.

    1976-01-01

    A description is given of a combined exhaust nozzle and intake pipe system to support a heat exchanger inside a nuclear reactor pressure vessel. It comprises a generally cylindrical part on the exhaust nozzle, the cylindrical part having an inside passage, a flange around the passage and provided with means to secure the exhaust nozzle to the reactor pressure vessel so as to make it fluidtight. The cylindrical part has an aperture inside to take the intake pipe inside the passage so as to enable the intake pipe to project into the heat exchanger. A collar made on the heat exchanger projects from the heat exchanger to the cylindrical nozzle component to establish communication with the inside passage for the fluid [fr

  9. Fundamental conceptual design of the experimental multi-purpose high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shimokawa, Junichi; Yasuno, Takehiko; Yasukawa, Shigeru; Mitake, Susumu; Miyamoto, Yoshiaki

    1975-06-01

    The fundamental conceptual design of the experimental multi-purpose very high-temperature gas-cooled reactor (experimental VHTR of thermal output 50 MW with reactor outlet-gas temperature 1,000 0 C) has been carried out to provide the operation modes of the system consisting of the reactor and the heat-utilization system, including characteristics and performance of the components and safety of the plant system. For the heat-utilization system of the plant, heat distribution, temperature condition, cooling system constitution, and the containment facility are specified. For the operation of plant, testing capability of the reactor and controlability of the system are taken into consideration. Detail design is made of the fuel element, reactor core, reactivity control and pressure vessel, and also the heat exchanger, steam reformer, steam generator, helium circulator, helium-gas turbine, and helium-gas purification, fuel handling, and engineered safety systems. Emphasis is placed on providing the increase of the reactor outlet-gas temperature. Fuel element design is directed to the prismatic graphite blocks of hexagonal cross-section accommodating the hollow or tubular fuel pins sheathed in graphite sleeve. The reactor core is composed of 73 fuel columns in 7 stages, concerning the reference design MK-II. Orificing is made in the upper portion of core; one orifice for every 7 fuel columns. Average core power density is 2.5 watts/cm 3 . Fuel temperature is kept below 1,300 0 C in rated power. The main components, i.e. pressure vessel, reformer, gas turbine and intermediate heat exchanger are designed in detail; the IHX is of a double-shell and helically-wound tube coils, the reformer is of a byonet tube type, and the turbine-compressor unit is of an axial flow type (turbine in 6 stages and compressor in 16 stages). (auth.)

  10. Outlet temperature measurement correction of Gd fuel assemblies at Dukovany NPP

    International Nuclear Information System (INIS)

    Jurickova, M.

    2008-01-01

    In year 2006 we started data processing from the Dukovany NPP operating history database that contained data from the old measurement system VK3 and the new Scorpio-VVER. The work has been done in cooperation with the reactor physicists at Dukovany NPP. Obtained data from database were compared with calculated parameters from 3D diffusion macrocode Mobydick. During the data processing it was found that the Gd fuel assemblies have different time plot of measured assembly outlet temperature compared to the non-Gd fuel assemblies. Experimental studies in RRC KI found that there is insufficient coolant mixing in the region from the fuel bundle to the fuel assembly thermocouple. Due to this fact the thermocouple measure temperature is systematically higher than real temperature. There are two methods to solve this problem. The first method analyses the flow and heat transfer in the region from the fuel bundle to the fuel assembly thermocouple - this method is developed in Skoda JS. The second method statistically studies differences between the measured and calculated temperature by the Mobydick code using the operational history database. Our study is focused on the second method. Several calculation methods for the correction of measured assembly outlet temperature were developed. All correction methods were applied to the measured temperatures from the Dukovany NPP operating history database and the methods were mutually compared. In near future it is planned to compare results of our chosen correction method with modeling method, which is developing in Skoda JS and it is planned to validate both of them. Consequently, the one of these correction methods will be implemented in the modernized Scorpio-VVER for Dukovany NPP. (author)

  11. Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Conlon, W.M.; Lahey, R.T. Jr.

    1982-12-01

    This report describes an experimental study of the influence of parallel channel effects (PCE) on the distribution of emergency core spray cooling water in a Boiling Water Nuclear Reactor (BWR) following a postulated design basis loss of coolant accident (LCA). The experiments were conducted in a scaled test section in which the reactor coolant was simulated by Freon-114 at conditions similar to those postulated to occur in the reactor vessel shortly after a LOCA. A BWR/4 was simulated by a (PCE) test section which contained three parallel heated channels to simulate fuel assemblies; a core bypass channel, and a jet pump channel. The test section also inlcuded scaled regions to simulate the lower and upper plena, downcomer, and steam separation regions of a BWR. A series of nine transient experiments were conducted, in which the lower plenum vaporization rate and heater rod power were varied while the core spray flow rate was held constant to simulate that of a BWR/4. During these experiments the flow distribution and heat transfer phenomena were observed and measured

  12. A probabilistic method for determining effluent temperature limits for flow instability for SRS reactors

    International Nuclear Information System (INIS)

    Hardy, B.J.; White, A.M.

    1990-06-01

    This manual describes the uncertainty analysis used to determine the effluent temperature limits for a Mark 22 charge in the Savannah River Site production reactors. The postulated accident scenario is a DEGB/LOCA resulting from a coolant pipe break at the plenum inlet accompanied by the safety rod failure described in the previous chapter. The analysis described in this manual is used to calculate the limits for the flow instability phase of the accident. For this phase of the accident, the limits criterion is that the Stanton number does not exceed 0.00455 [1]. The limits are determined for a specified 84% probability that the Stanton number will not exceed 0.00455 in any assembly in the core

  13. A probabilistic method for determining effluent temperature limits for flow instability for SRS reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hardy, B.J.; White, A.M.

    1990-06-01

    This manual describes the uncertainty analysis used to determine the effluent temperature limits for a Mark 22 charge in the Savannah River Site production reactors. The postulated accident scenario is a DEGB/LOCA resulting from a coolant pipe break at the plenum inlet accompanied by the safety rod failure described in the previous chapter. The analysis described in this manual is used to calculate the limits for the flow instability phase of the accident. For this phase of the accident, the limits criterion is that the Stanton number does not exceed 0.00455 [1]. The limits are determined for a specified 84% probability that the Stanton number will not exceed 0.00455 in any assembly in the core.

  14. A neutronics study for improving the safety and performance parameters of a 3600 MWth Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Sun, Kaichao; Krepel, Jiri; Mikityuk, Konstantin; Chawla, Rakesh

    2013-01-01

    Highlights: ► The potential for neutronics design optimization is assessed for a large SFR core. ► Both beginning-of-life and equilibrium fuel cycle conditions are considered. ► The sodium void effect is decomposed via a neutron balance based methodology. ► The optimized core options adopt an appropriate sodium plenum design to reduce the void effect. ► The introduction of moderator pins is considered for enhancing the Doppler effect. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many performance advantages, but has one dominating neutronics drawback – a positive sodium void reactivity. The starting point for the present study is an SFR core design considered in the Collaborative Project on the European Sodium-cooled Fast Reactor (CP-ESFR). The aim is to analyze, for this reference core, four safety and performance parameters from the viewpoint of four different optimization options, and to propose possible optimized core designs. In doing so, the study focuses not only on the beginning-of-life state of the core, but also on the beginning of equilibrium closed fuel cycle. The four studied optimization options are: (a) introducing an upper sodium plenum and boron layer, (b) varying the core height-to-diameter (H/D) ratio, (c) introducing moderator pins into the fuel assembly, and (d) modifying the initial plutonium content. The sensitivity of the void reactivity, Doppler constant, nominal reactivity and breeding gain has been evaluated. In particular, the void reactivity, which is the most crucial safety parameter for the SFR, has been decomposed into its reaction-wise, isotope-wise and energy-group-wise components using a methodology based on the neutron balance equation. Extended voiding in the upper sodium plenum region – in conjunction with the effect of a boron layer introduced above the plenum – is found to be particularly effective in the void effect reduction while, at the same time

  15. Summary of the experimental multi-purpose very high temperature gas cooled reactor design

    International Nuclear Information System (INIS)

    1984-12-01

    The report presents the design of Multi-purpose Very High Temperature Gas Cooled Reactor (the Experimental VHTR) based on the second stage of detailed design which was completed on March 1984, in the from of ''An application of reactor construction permit Appendix 8''. The Experimental VHTR is designed to satisfy with the design specification for the reactor thermal output 50 MW and reactor outlet temperature 950 0 C. The adequacy of the design is also checked by the safety analysis. The planning of plant system and safety is summarized such as safety design requirements and conformance with them, seismic design and plant arrangement. Concerning with the system of the Experimental VHTR the design basis, design data and components are described in the order. (author)

  16. Review of In-Service Inspection and Repair Technique Developments for French Liquid Metal Fast Reactors

    International Nuclear Information System (INIS)

    Baque, F.

    2005-01-01

    In-service monitoring of nuclear plants is indispensable for both the Operator and the Regulator. The notion of in-service monitoring ranges from the continuous monitoring of the reactor in operation to the thorough in-service reactor inspection during programmed shutdowns. However, the highly specific environment found in French liquid metal fast reactor plants - Phenix and Superphenix - makes monitoring and inspection complicated because of the use of a sodium coolant that is hot, opaque, and difficult to drain.The Commissariat a l'Energie Atomique, in collaboration with its traditional French partners, Electricite de France utilities and FRAMATOME/Novatome Engineering, decided to conduct a 6-yr research and development program (1994-2000) to explore this problem vis-a-vis Superphenix, as well as the possibilities of intervening within the reactor block or on components in a sodium environment. Furthermore, the safety reevaluation of Phenix, conducted between 1994 and 2003, represented an excellent 'test bench' during which the limits of inspection processes - applied to an integrated reactor concept - were surpassed using techniques such as fuel subassembly head scanning, ultrasonic examination of the core support, and visual inspection of the cover-gas plenum following a partial sodium draining. Repair techniques were investigated for cleaning of sodium wet structure surfaces, cutting of damaged parts, and welding in sodium aerosol atmosphere. Both conventional and laser processes were tested

  17. Prediction of thermal margin for external cooling of reactor vessel lower head during a severe accident

    International Nuclear Information System (INIS)

    Yoon, Ho Jun; Suh, Kune Y.

    1998-01-01

    In the TMI-2 accident, approximately nineteen (19) tons of molten core material drained into the lower plenum. One of the major findings from the TMI-2 Vessel Investigation Project was that one part of the reactor lower head wall estimated to have attained a temperature of 1100 .deg. C for about 30 minutes has seemingly experienced a comparatively rapid cooldown with no major threat to the vessel integrity. In this regard, recent empirical and analytical studies have shifted interests to such in-vessel retention designs or strategies as reactor cavity flooding, in-vessel flooding and engineered gap cooling of the vessel. Accurate thermohydrodynamic and creep deformation modeling and rupture prediction are the key to the success in developing practically useful in-vessel accident management strategies. As an advanced in-vessel design concept, the COrium Attak Syndrome Immunization Structures (COASIS) are being developed as prospective in-vessel retention devices for a next-generation LWR in concert with existing ex-vessel management measures. Both the engineered gap structures in -vessel (COASISI) and ex-vessel (COASISO) were demonstrated to maintain effective heat transfer geometry during molten core debris attack when applied to the TMI-2 and the Korean Standard Nuclear Power Plant (KSNPP) reactors. The likelihood of lower head creep rupture during a severe accident is found to be significantly suppressed by the COASIS options. In studying the in-vessel severe accident phenomena, one of the main goals is to verify the cooling mechanism in the reactor vessel lower plenum and thereby to prevent the vessel failure from thermal attack by the molten debris. This paper presents the first-principle calculation results for the thermal margin for the case of external cooling of the reactor vessel lower head. Adopting the method presented by F.B. Cheung, et al., we calculated the departure from nucleate boiling ratio (DNBR) for the three cases of pool boiling, flow boiling

  18. 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.

  19. Availability of healthier options in traditional and nontraditional rural fast-food outlets

    Directory of Open Access Journals (Sweden)

    McIntosh Alex

    2008-11-01

    Full Text Available Abstract Background Food prepared away from home has become increasingly popular to U.S. families, and may contribute to obesity. Sales have been dominated by fast food outlets, where meals are purchased for dining away from home or in the home. Although national chain affiliated fast-food outlets are considered the main source for fast food, fast foods are increasingly available in convenience stores and supermarkets/grocery stores. In rural areas, these nontraditional fast-food outlets may provide most of the opportunities for procurement of fast foods. Methods Using all traditional and nontraditio nal fast-food outlets identified in six counties in rural Texas, the type and number of regular and healthiermenu options were surveyed using on-site observation in all food venues that were primarily fast food, supermarket/grocery store, and convenience store and compared with 2005 Dietary Guidelines. Results Traditional fast-food outlets represented 84 (41% of the 205 opportunities for procurement of fast food; 109 (53.2% were convenience stores and 12 (5.8% supermarkets/grocery stores. Although a s imilar variety of regular breakfast and lunch/dinner entrées were available in traditional fast-food outlets and convenience stores, the variety of healthier breakfast and lunch/dinner entrées was significantly greater in fast food outlets. Compared with convenience stores, supermarkets/grocery stores provided a greater variety of regular and healthier entrées and lunch/dinner side dishes. Conclusion Convenience stores and supermarkets/grocery stores more than double the potential access to fast foods in this rural area than traditional fast-food outlets alone; however, traditional fast food outlets offer greater opportunity for healthier fast food options than convenience stores. A complete picture of fast food environment and the availability of healthier fast food options are essential to understand environmental influences on diet and health

  20. Study of the mechanisms for the emergency cooling of the core of the Radioisotope Producing Reator (RPR)

    International Nuclear Information System (INIS)

    Lacerda, F.C.

    1987-01-01

    The mechanisms for the emergency cooling of the core of the Radioisotope Producing Reactor (R.P.R.) are studied, in particular the thermal-hydraulic behaviour of the coolant after reactor shut-down. The coolant operates bd convection, and flows downward through the core passing into beel-shaped plenum that encloses the core and proceeding across the primary cooling loop. When the reactor is shut-down, the coolant flow undergoes a transient period until the steady state of natural convection is reached, after which the coolant flows upwards from the lower plenum. A plocking valve will be installed at the exit of the lower plenum, which will automatically shut in case of an accident that will involve the loss of flow in the primary circuit. The present work aims at evaluating the contribution of natural convection by natural recirculation in the core when the blocking valve is close, and via the external coolant circuit when the blocking valve is open. In particular, we study the natural self-regulating mechanisms of extraction of the heat generated by the fission product after reactor shut-down. (author) [pt

  1. Reactor hydrodynamics during the reflood phase of a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Gay, R.R.

    1977-01-01

    The thermohydraulics of a nuclear reactor during the reflood phase of a hypothetical loss-of-coolant accident can be represented by moving control volume methodology in which six control volumes are used to represent the downcomer, lower plenum, and reactor core. The one-dimensional, homogeneous, equilibrium constitutive equations for two-phase steam/water flow are solved in each control volume and connecting junctions. One of the three core control volumes represents the quench region; it changes size and position based on the axial location of the clad quench temperature and the condensed liquid level in the flow channel. The lengths of the remaining two core control volumes are determined by the position of the quench region. Simulation of actual reflood experiments demonstrates that the methodology predicts reflood-like flow oscillations and reproduces the correct trends in experimental data. The moving control volume methodology has proven itself as a valid concept for reflood hydrodynamics, but further development of the existing EFLOD code is required for simulation of actual reflood experiments

  2. Transient flow characteristics of nuclear reactor coolant pump in recessive cavitation transition process

    International Nuclear Information System (INIS)

    Wang Xiuli; Yuan Shouqi; Zhu Rongsheng; Yu Zhijun

    2013-01-01

    The numerical simulation calculation of the transient flow characteristics of nuclear reactor coolant pump in the recessive cavitation transition process in the nuclear reactor coolant pump impeller passage is conducted by CFX, and the transient flow characteristics of nuclear reactor coolant pump in the transition process from reducing the inlet pressure at cavitation-born conditions to NPSHc condition is studied and analyzed. The flow field analysis shows that, in the recessive cavitation transition process, the speed diversification at the inlet is relative to the bubble increasing, and makes the speed near the blade entrance increase when the bubble phase region becomes larger. The bubble generation and collapse will affect the the speed fluctuation near the entrance. The vorticity close to the blade entrance gradually increasing is influenced by the bubble phase, and the collapse of bubble generated by cavitation will reduce the vorticity from the collapse to impeller outlet. Pump asymmetric structure causes the asymmetry of the flow, velocity and outlet pressure distribution within every impeller flow passage, which cause the asymmetry of the transient radial force. From the dimensionless t/T = 0.6, the bubble phase starts to have impact on the impeller transient radial force, and results in the irregular fluctuations. (authors)

  3. Vertical pump with free floating check valve

    International Nuclear Information System (INIS)

    Lindsay, M.

    1980-01-01

    A vertical pump is described which has a bottom discharge with a free floating check valve disposed in the outlet plenum thereof. The free floating check valve comprises a spherical member with a hemispherical cage-like member attached thereto which is capable of allowing forward or reverse flow under appropriate conditions while preventing reverse flow under inappropriate conditions

  4. Cooling of pressurized water nuclear reactor vessels

    International Nuclear Information System (INIS)

    Curet, H.D.

    1978-01-01

    The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

  5. 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)

  6. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  7. Power source with spark-safe outlet

    Energy Technology Data Exchange (ETDEWEB)

    Tsesarenko, N P; Alekhin, A V

    1982-01-01

    The invention refers to the technique of electrical monitoring and control in systems operating in a spark-safe medium (for example, in coal mines). A more accurate area of application is mobile objects with autonomous source of electricity (mine diesel locomotives, battery electric locomotives etc.). The purpose of the invention is to simplify and to improve the reliability of the planned device, and also to expand the area of application for conditions when it is powered from an autonomous generator of direct voltage. This goal is achieved because the power source with spark-safe outlet (the source contains a thyristor of advance disconnection, connected by anode to the delimiting throttle, one outlet of which is connected to the capacitor included between the controlling electrode and the anode of the thyristor, and the capacitor is connected through the resistor parallel to the outlet clamps of the source, while the thyristor of emergency protection connected parallel to the inlet clamps of the power source) is additionally equipped with a current sensor, hercon, transistor key (included in series in the power circuit) and optron, whose emitter is connected parallel to the current sensor connected in series to the inlet of the power source, while the receiver of the optron is connected in a circuit for controlling the thyristor of emergency protection. Hercon is built into the core of the delimiting throttle and is connected to the circuit for controlling the transistor key.

  8. The Role at Rehabilitation in Treatment of Thoracic Outlet Syndrome

    Directory of Open Access Journals (Sweden)

    Mohammad Ali Hosseinian

    2003-01-01

    Full Text Available Objective: Thoracic outlet syndrome is a complex disorder caused by neurovascular irritation in the region of the thoracic outlet. The syndrome have been said to be mainly due to anomalous structures in the thoracic outlet, treatment for thoracic outlet syndrome varies among different institutions, and there has not been any standard program. In general conservative and surgical treatment can be do if necessary. Materials & Methods: The rehabilitation program consists of exercise and physiotherapy and brace designed to hold the posture in which thoracic outlet is enlarged. Exercise program was designed simple enough to be performed in the daily living or during work after minimal training and isometric exercises of Serratus anterior, Levator Scapulae and Erector Spinae muscles to be performed in one posture: flexion and elevation of scapular girdle and correction position of upper-thoracic spine. During 7 years, 131 cases of (T.O.S. were evaluated that 26 cases (20% have operated and 84 cases (64% have treated with conservative treatment and 21 cases (16% have been candidate for surgery but they didn't accepted. Results: All of the cases have treated with conservative treatment for four months. 84 cases responded well and no further treatment was needed. 47 cases were not satisfied with. The outcome of their treatment, that 26 cases have operated and 21 cases have not accepted the operation and continued the conservative treatment, they have had pain and slightly disability. 23 cases of operated group responded well and they have resumed to work, one case has had neuropraxia for about one year. Conclusion: Most cases of thoracic outlet syndrome (T.O.S. can be treated conservatively. Surgically treatment is indicated only in cases severe enough to make them disable to work. It is better all the patients undergo conservative treatment for at least four months then will decided for surgical treatment.

  9. Experiments and models of general corrosion and flow-assisted corrosion of materials in nuclear reactor environments

    Science.gov (United States)

    Cook, William Gordon

    Corrosion and material degradation issues are of concern to all industries. However, the nuclear power industry must conform to more stringent construction, fabrication and operational guidelines due to the perceived additional risk of operating with radioactive components. Thus corrosion and material integrity are of considerable concern for the operators of nuclear power plants and the bodies that govern their operations. In order to keep corrosion low and maintain adequate material integrity, knowledge of the processes that govern the material's breakdown and failure in a given environment are essential. The work presented here details the current understanding of the general corrosion of stainless steel and carbon steel in nuclear reactor primary heat transport systems (PHTS) and examines the mechanisms and possible mitigation techniques for flow-assisted corrosion (FAC) in CANDU outlet feeder pipes. Mechanistic models have been developed based on first principles and a 'solution-pores' mechanism of metal corrosion. The models predict corrosion rates and material transport in the PHTS of a pressurized water reactor (PWR) and the influence of electrochemistry on the corrosion and flow-assisted corrosion of carbon steel in the CANDU outlet feeders. In-situ probes, based on an electrical resistance technique, were developed to measure the real-time corrosion rate of reactor materials in high-temperature water. The probes were used to evaluate the effects of coolant pH and flow on FAC of carbon steel as well as demonstrate of the use of titanium dioxide as a coolant additive to mitigated FAC in CANDU outlet feeder pipes.

  10. Predictions on an HTR coolant composition after operational experience with experimental reactors

    International Nuclear Information System (INIS)

    Nieder, R.

    1981-01-01

    Long-term operational experience of the HTR experimental reactors Dragon (1966 - 1975), Peach Bottom (1967 - 1974) and AVR (since 1967) has yielded a large number of common quantitative and qualitative results about the sources and behaviour of helium impurities in the primary circuits. Additional information has also been obtained from experiments made at the three reactors. The results at the AVR are particularly interesting because the gas outlet temperature can be varied from 770 0 C to 950 0 C when the reactor power is kept constant. Hence they can be studied according to the temperature dependence of all chemical reactions. It should be possible to apply the results from the operating measurements and experiments made at the reactors, in particular the interrelation of the impurity concentrations, to future reactors. The absolute values of these impurity concentrations are obtained first and foremost by the corresponding helium purification constants

  11. Investigation of coolant mixture in pressurized water reactors at the Rossendorf mixing test facility ROCOM

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Richter, K.; Weiss, F.P.

    1999-01-01

    During the so-called boron dilution or cold water transients at pressurized water reactors too weakly borated water or too cold water, respectively, might enter the reactor core. This results in the insertion of positive reactivity and possibly leads to a power excursion. If the source of unborated or subcooled water is not located in all coolant loops but in selected ones only, the amount of reactivity insertion depends on the coolant mixing in the downcomer and lower plenum of the reactor pressure vessel (RPV). Such asymmetric disturbances of the coolant temperature or boron concentration might e.g. be the result of a failure of the chemical and volume control system (CVCS) or of a main steam line break (MSLB) that does only affect selected steam generators (SG). For the analysis of boron dilution or MSLB accidents coupled neutron kinetics/thermo-hydraulic system codes have been used. To take into account coolant mixing phenomena in these codes in a realistic manner, analytical mixing models might be included. These models must be simple and fast running on the one hand, but must well describe the real mixing conditions on the other hand. (orig.)

  12. Ventilation system in the RA reactor building - design specifications

    International Nuclear Information System (INIS)

    Badrljica, R.

    1984-09-01

    Protective role of the ventilation system of nuclear facilities involve construction of ventilation barriers which prevent release of radioactive particulates or gases, elimination od radioactive particulates and gases from the air which is released from contaminated zones into the reactor environment. Ventilation barriers are created by dividing the building into a number of ventilation zones with different sub pressure compared to the atmospheric pressure. The RA reactor building is divided into four ventilation zones. First zone is the zone of highest risk. It includes reactor core with horizontal experimental channels, underground rooms of the primary coolant system (D 2 O), helium system, hot cells and the space above the the reactor core. Second zone is the reactor hall and the room for irradiated fuel storage. The third zone includes corridors in the basement, ground floor and first floor where the probability of contamination is small. The fourth zone includes the annex where the contamination risk is low. There is no have natural air circulation in the reactor building. Ventilators for air input and outlet maintain the sub pressure in the building (pressure lower than the atmospheric pressure). This prevents release of radioactivity into the atmosphere [sr

  13. WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

    International Nuclear Information System (INIS)

    Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

    1982-01-01

    At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru

  14. Engineering Design of a Double Reactor for Spent Fuel Oxidation

    International Nuclear Information System (INIS)

    Kim, Young-Hwan; Lee, Jae-Won; Lee, Ju-Ho; Cho, Yung-Zun; Ahn, Do-Hee

    2015-01-01

    In this study, for a performance enhancement of the oxidation treatment device recovery ratio, the first performance test of the existing device (prototype) oxidation treatment device was carried out. In addition, by analyzing the result, the size of the reactor with a 1 kg HM/batch for a recovery ratio enhancement was decided, and the structure of the reactor was derived as a double structure reactor with a mesh type drum. The principle and structure of this device are as follows. The pellet of the supplied rods is oxidized in 500 .deg. C reactor A, and penetrates reactor B to form a uniform powder. In addition, if it is rotated in the reverse direction, the powder and hull are separated. The device is composed of a reactor module, driving module, heater module, support module, outlet module, etc. In addition, by reflecting the enhancements, a voloxidizer with a double reactor was designed and manufactured, and a second performance test was carried out. Using a 30 mm hull and simulated powders (balls), as a result of carrying out the enhanced device performance test, the hull recovery ratio was 100%, and the simulated powder recovery ratio was 99% or more

  15. Engineering Design of a Double Reactor for Spent Fuel Oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young-Hwan; Lee, Jae-Won; Lee, Ju-Ho; Cho, Yung-Zun; Ahn, Do-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    In this study, for a performance enhancement of the oxidation treatment device recovery ratio, the first performance test of the existing device (prototype) oxidation treatment device was carried out. In addition, by analyzing the result, the size of the reactor with a 1 kg HM/batch for a recovery ratio enhancement was decided, and the structure of the reactor was derived as a double structure reactor with a mesh type drum. The principle and structure of this device are as follows. The pellet of the supplied rods is oxidized in 500 .deg. C reactor A, and penetrates reactor B to form a uniform powder. In addition, if it is rotated in the reverse direction, the powder and hull are separated. The device is composed of a reactor module, driving module, heater module, support module, outlet module, etc. In addition, by reflecting the enhancements, a voloxidizer with a double reactor was designed and manufactured, and a second performance test was carried out. Using a 30 mm hull and simulated powders (balls), as a result of carrying out the enhanced device performance test, the hull recovery ratio was 100%, and the simulated powder recovery ratio was 99% or more.

  16. Female outlet obstruction constipation: assessment with MR defecography

    International Nuclear Information System (INIS)

    Li Min; Jiang Tao; Yang Xinqing; Peng Peng; Wang Wenchuan

    2010-01-01

    Objective: Using MR defecography to assess the morphological and functional anorectal anomalies related to female outlet obstruction constipation, and evaluate the joint disease of' anterior and mid pelvic. Methods: One hundred and seven female patients, aged 20 to 84 years (average, 55 years), were diagnosed as outlet obstruction constipation based on clinical symptoms and signs. They all received MR defecography in our institution. The high compliance homemade balloon was inserted into rectum to simulate stool. Then relevant measurements were obtained during rest, squeezing and straining, respectively. Results: In all the 107 cases, 70 (65.4%) presented rectocele on dynamic MRI; 28 (26.2%) presented anismus; 60 (56.1%) presented cystocele; 59 presented vaginal or cervical prolapse(55.1%); and, 54 (50.5%) presented descending perineum. In 85 females (79.4%) multiple disorders were detected, involving more than one pelvic compartment. Conclusion: MR defecography allowed to accurately evaluate the morphological and functional anorectal anomalies related to female outlet obstruction constipation, and the joint disease of anterior and mid pelvic. (authors)

  17. Availability of healthier options in traditional and nontraditional rural fast-food outlets

    OpenAIRE

    Creel, Jennifer S; Sharkey, Joseph R; McIntosh, Alex; Anding, Jenna; Huber, J Charles

    2008-01-01

    Abstract Background Food prepared away from home has become increasingly popular to U.S. families, and may contribute to obesity. Sales have been dominated by fast food outlets, where meals are purchased for dining away from home or in the home. Although national chain affiliated fast-food outlets are considered the main source for fast food, fast foods are increasingly available in convenience stores and supermarkets/grocery stores. In rural areas, these nontraditional fast-food outlets may ...

  18. Design of a reactor inlet temperature controller for EBR-2 using state feedback

    International Nuclear Information System (INIS)

    Vilim, R.B.; Planchon, H.P.

    1990-01-01

    A new reactor inlet temperature controller for pool type liquid-metal reactors has been developed and will be tested in EBR-II. The controller makes use of modern control techniques to take into account stratification and mixing in the cold pool during normal operation. Secondary flowrate is varied so that the reactor inlet temperature tracks a setpoint while reactor outlet temperature, primary flowrate and secondary cold leg temperature are treated as exogenous disturbances and are free to vary. A disturbance rejection technique minimizes the effect of these disturbances on inlet temperature. A linear quadratic regulator improves inlet temperature response. Tests in EBR-II will provide experimental data for assessing the performance improvements that modern control can produce over the existing EBR-II analog inlet temperature controller. 10 refs., 8 figs

  19. The Liquid Annular Reactor System (LARS) propulsion

    International Nuclear Information System (INIS)

    Powell, J.; Ludewig, H.; Horn, F.; Lenard, R.

    1990-01-01

    A concept for very high specific impulse (greater than 2000 seconds) direct nuclear propulsion is described. The concept, termed the liquid annular reactor system (LARS), uses liquid nuclear fuel elements to heat hydrogen propellant to very high temperatures (approximately 6000 K). Operating pressure is moderate (approximately 10 atm), with the result that the outlet hydrogen is virtually 100 percent dissociated to monatomic H. The molten fuel is contained in a solid container of its own material, which is rotated to stabilize the liquid layer by centripetal force. LARS reactor designs are described, together with neutronic and thermal-hydraulic analyses. Power levels are on the order of 200 megawatts. Typically, LARS designs use seven rotating fuel elements, are beryllium moderated, and have critical radii of approximately 100 cm (core L/D approximately equal to 1.5)

  20. Impact of remodeling and rehabilitation of irrigation outlets on water distribution of a canal in Punjab, Pakistan

    International Nuclear Information System (INIS)

    Bodla, H.; Latif, M.

    2009-01-01

    The study was undertaken to investigate water distribution along a distributary canal located in the southern part of the Punjab Province. It is a large size distributary having 353 cusecs of authorized discharge. This distributary was subjected to a series of problems including but not limited to (i) withdrawal of water by illegal means, (ii) design and construction flaws in the outlets, (iii) improper selection of the type of outlets and many others. The outlets were intentionally designed wrongly by using fictitious hydraulics data to provide undue benefits to the irrigators. During construction, setting of the outlets was also intentionally fixed at lower level than the designed. Investigations were carried out to evaluate hydraulic performance of all the outlets of the channel. Based on the observed data capacity statements of all the outlets were revised. The outlets were redesigned on the basis of actual hydraulic data of each outlet. Most of the non-modular outlets (Pipe and Scratchley) were converted to semi-modular outlets (OFRB and APM). With implementation of new and modified design of the outlets at the site, equity of water distribution has been improved. The results revealed that design of the outlets had a significant impact on equitable distribution of water along the distribution. (author)

  1. Nuclear design for high temperature gas cooled reactor (GTHTR300C) using MOX fuel

    International Nuclear Information System (INIS)

    Mouri, Tomoaki; Kunitomi, Kazuhiko

    2008-01-01

    A design study of the hydrogen cogeneration high temperature gas cooled reactor (GTHTR300C) that can produce both electricity and hydrogen has been carried out in Japan Atomic Energy Agency. The GTHTR300C is the system with thermal power of 600MW and reactor outlet temperature of 950degC, which is expected to supply the hydrogen to fuel cell vehicles after 2020s. In future, the full deployment of fast reactor cycle without natural uranium will demand the use of Mixed-Oxide (MOX) fuels in the GTHTR300C. Therefore, a nuclear design was performed to confirm the feasibility of the reactor core using MOX fuels. The designed reactor core has high performance and meets safety requirements. In this paper, the outline of the GTHTR300C and the nuclear design of the reactor core using MOX fuels are described. (author)

  2. Benign Strictures of the Esophagus and Gastric Outlet: Interventional Management

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jin Hyoung; Shin, Ji Hoon; Song, Ho Young [University of Ulsan College of Medicine, Asan Medical Center, Seoul (Korea, Republic of)

    2010-10-15

    Benign strictures of the esophagus and gastric outlet are difficult to manage conservatively and they usually require intervention to relieve dysphagia or to treat the stricture-related complications. In this article, authors review the non-surgical options that are used to treat benign strictures of the esophagus and gastric outlet, including balloon dilation, temporary stent placement, intralesional steroid injection and incisional therapy

  3. Using public health and community partnerships to reduce density of alcohol outlets.

    Science.gov (United States)

    Jernigan, David H; Sparks, Michael; Yang, Evelyn; Schwartz, Randy

    2013-04-11

    Excessive alcohol use causes approximately 80,000 deaths in the United States each year. The Guide to Community Preventive Services recommends reducing the density of alcohol outlets - the number of physical locations in which alcoholic beverages are available for purchase either per area or per population - through the use of regulatory authority as an effective strategy for reducing excessive alcohol consumption and related harms. We briefly review the research on density of alcohol outlets and public health and describe the powers localities have to influence alcohol outlet density. We summarize Regulating Alcohol Outlet Density: An Action Guide, which describes steps that local communities can take to reduce outlet density and the key competencies and resources of state and local health departments. These include expertise in public health surveillance and evaluation methods, identification and tracking of outcome measures, geographic information systems (GIS) mapping, community planning and development of multisector efforts, and education of community leaders and policy makers. We illustrate the potential for partnerships between public health agencies and local communities by presenting a contemporary case study from Omaha, Nebraska. Public health agencies have a vital and necessary role to play in efforts to reduce alcohol outlet density. They are often unaware of the potential of this strategy and have strong potential partners in the thousands of community coalitions nationwide that are focused on reducing alcohol-related problems.

  4. Validation of presence of supermarkets and fast-food outlets in Copenhagen

    DEFF Research Database (Denmark)

    Svastisalee, Chalida M; Holstein, Bjørn E; Due, Pernille

    2012-01-01

    We examined the quality of food outlet addresses provided by secondary sources and determined whether they could be physically located in the field.......We examined the quality of food outlet addresses provided by secondary sources and determined whether they could be physically located in the field....

  5. Application of radcal gamma thermometer assemblies for coolant monitoring in Ringhals W-PWRs

    International Nuclear Information System (INIS)

    Smith, R.D.; Romslo, K.; Moen, Oe.

    1982-07-01

    A study has been carried out investigating how Radcal Gamma Thermometers (RGTs) can be used for coolant inventory and core cooling monitoring in the Ringhals Westinghouse PWRs. The study concludes that two types of RGT rods would be required to come up with a complete solution covering both coolant inventory and core cooling monitoring. Above-core RGT rods will be installed in the guide tubes housing the outlet thermocouples. The Above-Core RGT rod is designed with 8 sensors where 4 are located in the upper head and 4 in the plenum. This rod will give an early warning about loss of coolant or void formation in the space from top of fuel to the reactor lid. A ninth thermocouple in this rod will measure the core outlet temperature as did the thermocouple the RGT rod replaced. The Above-Core RGT rods will give an early warning about approach to Inadequate Core Cooling (ICC) by measuring the collapsed water level inside the thermocouple guide tube. Four such rods are recommended per reactor. In-Core RGT rods are inserted from the seal table. These rods will give the information required for intelligent accident management in case ICC has developed. The signals obtainable from the rods will give direct information about fuel decay heat, core heat transfer conditions, core temperature and core coolant water level. The In-Core RGT rods can be used for local power monitoring during normal operation. Such a system can be shown to be economically motivated from a reactor operation point of view due to increased sensor lifetime, more accurate local power measurements, simpler physics corrections to signals, lower exposure to maintenance personnel. The signal transmission to the control room has been discussed, and ways have been indicated for presenting the information available to the operators. (Authors)

  6. Beyond Supermarkets: Food Outlet Location Selection in Four U.S. Cities Over Time.

    Science.gov (United States)

    Rummo, Pasquale E; Guilkey, David K; Ng, Shu Wen; Popkin, Barry M; Evenson, Kelly R; Gordon-Larsen, Penny

    2017-03-01

    Understanding what influences where food outlets locate is important for mitigating disparities in access to healthy food outlets. However, few studies have examined how neighborhood characteristics influence the neighborhood food environment over time, and whether these relationships differ by neighborhood-level income. Neighborhood-level data from four U.S. cities (Birmingham, AL; Chicago, IL; Minneapolis, MN; Oakland, CA) from 1986, 1993, 1996, 2001, 2006, and 2011 were used with two-step econometric models to estimate longitudinal associations between neighborhood-level characteristics (z-scores) and the log-transformed count/km 2 (density) of food outlets within real estate-derived neighborhoods. Associations were examined with lagged neighborhood-level sociodemographics and lagged density of food outlets, with interaction terms for neighborhood-level income. Data were analyzed in 2016. Neighborhood-level income at earlier years was negatively associated with the current density of convenience stores (β= -0.27, 95% CI= -0.16, -0.38, prestaurant density in low-income neighborhoods (10th percentile of income: β= -0.17, 95% CI= -0.34, -0.002, p=0.05), and the density of smaller grocery stores across all income levels (β= -0.27, 95% CI= -0.45, -0.09, p=0.003). There was a lack of policy-relevant associations between the pre-existing food environment and the current density of food outlet types, including supermarkets. Socioeconomically disadvantaged and minority populations may attract "unhealthy" food outlets over time. To support equal access to healthy food outlets, the availability of "less healthy" food outlets types may be relatively more important than the potential lack of supermarkets or full-service restaurants. Copyright © 2016 American Journal of Preventive Medicine. Published by Elsevier Inc. All rights reserved.

  7. Improvements to the sodium supply system of a nuclear reactor core

    International Nuclear Information System (INIS)

    Chevallier, Rene; Marchais, Christian.

    1981-01-01

    This invention concerns an improvement to the sodium supply system of a nuclear reactor core and, in particular, concerns the area included between the outlet of the primary circulation pumps and the core proper. A simplified structure and a lightening of all this linking area between the circulation pumps and the distribution tank under the core is achieved and this results in a very significant reduction in the risks of deterioration and in a definite increase in the reliability of the reactor. The invention is therefore an improvement to the sodium supply system of the nuclear reactor core vessel with incorporated exchangers, in which the cool sodium, after passing through the primary exchangers, is collected in a ring compartment from whence it is taken up by the pumps and moved to at least one pipe reaching a distribution tank located under the reactor core [fr

  8. Method and apparatus for removing radioactive gases from a nuclear reactor

    International Nuclear Information System (INIS)

    Frumerman, R.; Brown, W.W.

    1975-01-01

    A description is given of a method for removing radioactive gases from a nuclear reactor including the steps of draining coolant from a nuclear reactor to a level just below the coolant inlet and outlet nozzles to form a vapor space and then charging the space with an inert gas, circulating coolant through the reactor to assist the release of radioactive gases from the coolant into the vapor space, withdrawing the radioactive gases from the vapor space by a vacuum pump which then condenses and separates water from gases carried forward by the vacuum pump, discharging the water to a storage tank and supplying the separated gases to a gas compressor which pumps the gases to gas decay tanks. After the gases in the decay tanks lose their radioactive characteristics, the gases may be discharged to the atmosphere or returned to the reactor for further use

  9. Study on partial overheat of the isolated phase busbar outlet box in Qinshan NPP phase Ⅱ

    International Nuclear Information System (INIS)

    Tang Fangxuan; Zhang Jian; Zeng Limin; Bao Yanxing; Zhang Lie; Yang Yuemin

    2013-01-01

    This paper recommended the structure of the isolated phase busbar outlet box installed in Qinshan II. The study on partial overheat of the outlet box shows that the ultimate causes are the loss of concentrated eddy current and short of cooling. So the improvement principles of 'distributing eddy current, cutting off inductive circle current and strengthening of ventilation' were determined. A new structure test outlet box was designed and manufactured, and the temperature rising experiment was carried out. Some alterations were made in the new structure outlet box, e.g. isolating materials were added between side plates of the upper outlet box, and also between the upper and lower outlet box. Two cooling blowers were added to the upper outlet box. After putting into operation, the hot-spot temperature of the new outlet box was greatly lowered down. Thus the operation environment was improved, and the operation safety ensured. It can be useful references for analyzing and dealing with similar problems. (authors)

  10. Spatial accessibility to physical activity facilities and to food outlets and overweight in French youth.

    Science.gov (United States)

    Casey, R; Chaix, B; Weber, C; Schweitzer, B; Charreire, H; Salze, P; Badariotti, D; Banos, A; Oppert, J-M; Simon, C

    2012-07-01

    Some characteristics of the built environment have been associated with obesity in youth. Our aim was to determine whether individual and environmental socio-economic characteristics modulate the relation between youth overweight and spatial accessibility to physical activity (PA) facilities and to food outlets. Cross-sectional study. 3293 students, aged 12 ± 0.6 years, randomly selected from eastern France middle schools. Using geographical information systems (GIS), spatial accessibility to PA facilities (urban and nature) was assessed using the distance to PA facilities at the municipality level; spatial accessibility to food outlets (general food outlets, bakeries and fast-food outlets) was calculated at individual level using the student home address and the food outlets addresses. Relations of weight status with spatial accessibility to PA facilities and to food outlets were analysed using mixed logistic models, testing potential direct and interaction effects of individual and environmental socio-economic characteristics. Individual socio-economic status modulated the relation between spatial accessibility to PA facilities and to general food outlets and overweight. The likelihood of being overweight was higher when spatial accessibility to urban PA facilities and to general food outlets was low, but in children of blue-collar-workers only. The odds ratio (OR) (95% confidence interval) for being overweight of blue-collar-workers children compared with non-blue-collar-workers children was 1.76 (1.25-2.49) when spatial accessibility to urban PA facilities was low. This OR was 1.86 (1.20-2.86) when spatial accessibility to general food outlets was low. There was no significant relationship of overweight with either nature PA facilities or other food outlets (bakeries and fast-food outlets). These results indicate that disparities in spatial accessibility to PA facilities and to general food outlets may amplify the risk of overweight in socio

  11. Data on loss of off-site electric power simulation tests of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

    2002-07-01

    The high temperature engineering test reactor (HTTR), the first high temperature gas-cooled reactor (HTGR) in Japan, achieved the first full power of 30 MW on December 7 in 2001. In the rise-to-power test of the HTTR, simulation tests on loss of off-site electric power from 15 and 30 MW operations were carried out by manual shutdown of off-site electric power. Because helium circulators and water pumps coasted down immediately after the loss of off-site electric power, flow rates of helium and water decreased to the scram points. To shut down the reactor safely, the subcriticality should be kept by the insertion of control rods and the auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components. About 50 s later from the loss of off-site electric power, the auxiliary cooling system started up by supplying electricity from emergency power feeders. Temperature of hot plenum block among core graphite structures decreased continuously after the startup of the auxiliary cooling system. This report describes sequences of dynamic components and transient behaviors of the reactor and its cooling system during the simulation tests from 15 and 30 MW operations. (author)

  12. Potentialities of high temperature reactors (HTR)

    International Nuclear Information System (INIS)

    Hittner, D.

    2001-01-01

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

  13. Design of high temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Saito, Shinzo; Tanaka, Toshiyuki; Sudo, Yukio

    1994-09-01

    Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 850degC for rated operation and 950degC for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests. JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R and D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation. (author)

  14. The Preliminary Study of High Temperature Gas Cooled Reactors (HTGRs) Technology

    International Nuclear Information System (INIS)

    Nurfarhana Ayuni Joha; Izhar Abu Hussin; Ridzuan Abdul Mutalib

    2015-01-01

    High Temperature Gas Cooled Reactors (HTGRs) have attracted worldwide interest because of their high outlet temperatures, which allow them to be used for applications beyond electricity generation. HTGRs have been built and operated since as far back as the 1970s. Experimental and demonstration reactors of this type have operated in China, Great Britain, Germany, Japan, and the United States of America. This paper is written to share the valuable knowledge and information of HTGRs technology as a mean to enrich peoples understanding of the technology. This paper will present the technological features of HTGRs that allow for a modular design with inherently safe characteristics. (author)

  15. Safety design/analysis and scenario for prevention of CDA with ECCS in lead-bismuth-cooled fast reactor

    International Nuclear Information System (INIS)

    Minoru, Takahashi; Vaclav, Dostal; Abu Khalid, Rivai; Novitrian; Yumi, Yamada

    2007-01-01

    Safety design has been developed to show safety feature of Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR). The core is designed to have negative void reactivity even if the entire core and upper plenum are voided by steam intrusion from above. In-vessel type control rod driving mechanisms are used to prevent control rods from accidental ejection due to high pressure in the reactor vessel. In cases of coolant leakage from reactor vessel and feed water pipes, Pb-Bi coolant level in the reactor vessel is kept at the required level for decay heat removal by means of closed type guard vessel. Dual pipes are adopted to avoid leak of water in the feedwater system. Pump trip in feedwater systems initiates loss of coolant flow (LOF) event, although there is no concern of loss of flow accident due to primary pump trip. Injection of high pressure water slows down the flow-coast-down of feedwater at the LOF event. It has been evaluated that the fuel temperature is kept lower than safety limits at the unprotected loss of flow and heat sink (ATWS). A scenario for prevention of the core disruptive accident (CDA) with the emergency core cooling system (ECCS) is examined. The reactor becomes super-critical when the reactor vessel is filled with water. It is necessary to use water with boric acid for the ECC system, and additional backup rods for sub-critical core in water injection. (authors)

  16. 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.

  17. The prospects of making small retail outlets in the Townships aggressively competitive

    Directory of Open Access Journals (Sweden)

    Malefane Johannes Lebusa

    2013-12-01

    Full Text Available Historically, township Small Retail Outlets were mostly established for survival and operated under a generally closed market system where the competition was not very strong. However, with the advent of democracy many people lost their formal income through retrenchments and out of desperation, many of these people opened Small Retail Outlets thus most of the existing and new entrants into the township market were unskilled or semiskilled labourers with little or no formal skills in business or entrepreneurship. Such efforts were rarely guided by any specific and informed strategy of identifying and exploiting a gap in the market. With the consolidation of the free market system under democracy, big brand businesses such as Shoprite Checkers and Small Retail Outlets of foreign nationals with different strategies entered and competed in this township market. With fewer formal skills in business and entrepreneurship, the owners of the Small Retail Outlets struggled to compete and thrive under these relatively new economic conditions. Given this situation, I conducted semi-structured interviews with fifteen of these traditional Small Retail Outlets to find out and better understand the challenges they face and the skills that might be needed to aggressively compete in this space. Based on these findings and understandings, I further examined these issues and suggest infusions of specific entrepreneurship skills that could develop their aggressive competitiveness. Keywords: entrepreneurship, competitiveness, small retail outlets, shopping complexes, innovation

  18. Elemental mercury vapor capture by powdered activated carbon in a fluidized bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fabrizio Scala; Riccardo Chirone; Amedeo Lancia [Istituto di Ricerche sulla Combustione - CNR, Napoli (Italy)

    2011-06-15

    A bubbling fluidized bed of inert material was used to increase the activated carbon residence time in the reaction zone and to improve its performance for mercury vapor capture. Elemental mercury capture experiments were conducted at 100{sup o}C in a purposely designed 65 mm ID lab-scale pyrex reactor, that could be operated both in the fluidized bed and in the entrained bed configurations. Commercial powdered activated carbon was pneumatically injected in the reactor and mercury concentration at the outlet was monitored continuously. Experiments were carried out at different inert particle sizes, bed masses, fluidization velocities and carbon feed rates. Experimental results showed that the presence of a bubbling fluidized bed led to an increase of the mercury capture efficiency and, in turn, of the activated carbon utilization. This was explained by the enhanced activated carbon loading and gas-solid contact time that establishes in the reaction zone, because of the large surface area available for activated carbon adhesion/deposition in the fluidized bed. Transient mercury concentration profiles at the bed outlet during the runs were used to discriminate between the controlling phenomena in the process. Experimental data have been analyzed in the light of a phenomenological framework that takes into account the presence of both free and adhered carbon in the reactor as well as mercury saturation of the adsorbent. 14 refs., 7 figs.

  19. The association of alcohol outlet density with illegal underage adolescent purchasing of alcohol.

    Science.gov (United States)

    Rowland, Bosco; Toumbourou, John W; Livingston, Michael

    2015-02-01

    Although previous studies have suggested that greater community densities of alcohol sales outlets are associated with greater alcohol use and problems, the mechanisms are unclear. The present study examined whether density was associated with increased purchasing of alcohol by adolescents younger than the legal purchase age of 18 in Australia. The number of alcohol outlets per 10,000 population was identified within geographic regions in Victoria, Australia. A state-representative student survey (N = 10,143) identified adolescent reports of purchasing alcohol, and multilevel modeling was then used to predict the effects for different densities of outlet types (packaged, club, on-premise, general, and overall). Each extra sales outlet per 10,000 population was associated with a significant increase in the risk of underage adolescent purchasing. The strongest effect was for club density (odds ratio = 1.22) and packaged (takeaway) outlet density (odds ratio = 1.12). Males, older children, smokers, and those with substance-using friends were more likely to purchase alcohol. One mechanism by which alcohol sales outlet density may influence population rates of alcohol use and related problems is through increasing the illegal underage purchasing of alcohol. Copyright © 2015 Society for Adolescent Health and Medicine. Published by Elsevier Inc. All rights reserved.

  20. 16 CFR Appendix B to Part 436 - Sample Item 20(1) Table-Systemwide Outlet Summary

    Science.gov (United States)

    2010-01-01

    ... DISCLOSURE REQUIREMENTS AND PROHIBITIONS CONCERNING FRANCHISING Pt. 436, App. B Appendix B to Part 436—Sample... 1Outlet Type Column 2Year Column 3Outlets at the Start of the Year Column 4Outlets at the End of the Year...

  1. Methodological Approaches to Locating Outlets of the Franchise Retail Network

    OpenAIRE

    Grygorenko Tetyana M.

    2016-01-01

    Methodical approaches to selecting strategic areas of managing the future location of franchise retail network outlets are presented. The main stages in the assessment of strategic areas of managing the future location of franchise retail network outlets have been determined and the evaluation criteria have been suggested. Since such selection requires consideration of a variety of indicators and directions of the assessment, the author proposes a scale of evaluation, which ...

  2. Nodalization qualification process of the PSBVVER facility for the Cathare2 thermal-hydraulic code

    International Nuclear Information System (INIS)

    Del Nevo, A.; Araneo, D.; D'Auria, F.; Galassi, G.

    2004-01-01

    The present document deals with the nodalization qualification process of the PSB-VVER test facility for Cathare2 code. PSB-VVER facility is a 1/300 volume scale model of a VVER-1000, reactor installed at Electrogorsk Research and Engineering Centre in 1998. The version V1.5b of the Cathare2 code has been used. In order to evaluate the nodalization performance, the qualifying procedure set up at the DIMNP of Pisa University (UNIPI) has been applied that foresees two qualification levels: a 'steady state' level and an 'on transient' level. After the steady state behavior check of the nodalization, it has been preformed the on transient qualification the PSB-VVER test 2. It is a 11% equivalent break in Upper Plenum with the actuation of one high pressure injection system, connected to the hot leg of the loop 4, and 4 passive systems (ECCS hydro-accumulators), connected to the outlet plenum and to the inlet chamber of the downcomer. The low-pressure injection system is not available in the test. The goal of this paper is to demonstrate that the first step of the nodalization qualification adopted for the PSB test analyses is achieved and the PSB facility input deck is available and ready to use. The quantitative accuracy of the performed calculation has been evaluated by using the FFT-BM tool developed at the University of Pisa.(author)

  3. Structure optimization of CFB reactor for moderate temperature FGD

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yuan; Zhang, Jie; Zheng, Kai; You, Changfu [Tsinghua Univ., Beijing (China). Dept. of Thermal Engineering; Ministry of Education, Beijing (China). Key Lab. for Thermal Science and Power Engineering

    2013-07-01

    The gas velocity distribution, sorbent particle concentration distribution and particle residence time in circulating fluidized bed (CFB) reactors for moderate temperature flue gas desulfurization (FGD) have significant influence on the desulfurization efficiency and the sorbent calcium conversion ratio for sulfur reaction. Experimental and numerical methods were used to investigate the influence of the key reactor structures, including the reactor outlet structure, internal structure, feed port and circulating port, on the gas velocity distribution, sorbent particle concentration distribution and particle residence time. Experimental results showed that the desulfurization efficiency increased 5-10% when the internal structure was added in the CFB reactor. Numerical analysis results showed that the particle residence time of the feed particles with the average diameter of 89 and 9 {mu}m increased 40% and 17% respectively, and the particle residence time of the circulating particles with the average diameter of 116 {mu}m increased 28% after reactor structure optimization. The particle concentration distribution also improved significantly, which was good for improving the contact efficiency between the sorbent particles and SO{sub 2}. In addition, the optimization guidelines were proposed to further increase the desulfurization efficiency and the sorbent calcium conversion ratio.

  4. Plant accident dynamics of high-temperature reactors with direct gas turbine cycle

    International Nuclear Information System (INIS)

    Waloch, M.L.

    1977-01-01

    In the paper submitted, a one-dimensional accident simulation model for high-temperature reactors with direct-cycle gas turbine (single-cycle facilities) is described. The paper assesses the sudden failure of a gas duct caused by the double-ended break of one out of several parallel pipes before and behind the reactor for a non-integrated plant, leading to major loads in the reactor region, as well as the complete loss of vanes of the compressor for an integrated plant. The results of the calculations show especially high loads for the break of a hot-gas pipe immediately behind the flow restrictors of the reactor outlet, because of prolonged effects of pressure gradients in the reactor region and the maximum core differential pressure. A plant accident dynamics calculation therefore allows to find a compromise between the requirements of stable compressor operation, on the one hand, and small loads in the reactor in the course of an accident, on the other, by establishing in a co-ordinated manner the narrowing ratio of the flow restrictors. (GL) [de

  5. Investigation of analytical methods in thermal stratification analysis. Evaluation of flow rates through flow holes for normal and scram conditions of 40% power operation with AQUA code

    International Nuclear Information System (INIS)

    Doi, Yoshihiro; Muramatsu, Toshiharu

    1997-08-01

    Thermal stratification phenomena are observed in an upper plenum of liquid metal fast breeder reactors (LMFBRs) under reactor scram conditions, which give rise to thermal stress on structural components. Therefore it is important to evaluate characteristics of phenomena in the design of the internal structure in an LMFBR plenum. To evaluate flow rates through flow holes of the prototype fast breeder reactor, MONJU, numerical analyses were carried out with AQUA code for normal and scram conditions with 40% power operation. Through comparison of analysis results and measured temperature, thermal stratification phenomena in 300 second period after the scram was evaluated. Flow rate through the upper flow holes, the lower flow holes and annular gap between the inner barrel and the reactor vessel were evaluated with the measured temperature and the analysis results individually. (J.P.N.)

  6. Judgement on the data for fuel assembly outlet temperatures of WWER fuel assemblies in power reactors based on measurements with experimental fuel assemblies

    International Nuclear Information System (INIS)

    Krause, F.

    1986-01-01

    In the period from 1980 to 1985, in the Rheinsberg nuclear power plant experimental fuel assemblies were used on lattices at the periphery of the core. These particular fuel assemblies dispose of an extensive in-core instrumentation with different sensors. Besides this, they are fit out with a device to systematically thottle the coolant flow. The large power gradient present at the core position of the experimental fuel assembly causes a temperature profile along the fuel assemblies which is well provable at the measuring points of the outlet temperature. Along the direction of flow this temperature profile in the coolant degrades only slowly. This effect is to be taken into account when measuring the fuel assembly outlet temperature of WWER fuel assemblies. Besides this, the results of the measurements hinted both at a γ-heating of the temperature measuring points and at tolerances in the calculation of the micro power density distribution. (author)

  7. Heat removal performance of auxiliary cooling system for the high temperature engineering test reactor during scrams

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Tachibana, Yukio; Iyoku, Tatsuo; Takenaka, Satsuki

    2003-01-01

    The auxiliary cooling system of the high temperature engineering test reactor (HTTR) is employed for heat removal as an engineered safety feature when the reactor scrams in an accident when forced circulation can cool the core. The HTTR is the first high temperature gas-cooled reactor in Japan with reactor outlet gas temperature of 950 degree sign C and thermal power of 30 MW. The auxiliary cooling system should cool the core continuously avoiding excessive cold shock to core graphite components and water boiling of itself. Simulation tests on manual trip from 9 MW operation and on loss of off-site electric power from 15 MW operation were carried out in the rise-to-power test up to 20 MW of the HTTR. Heat removal characteristics of the auxiliary cooling system were examined by the tests. Empirical correlations of overall heat transfer coefficients were acquired for a helium/water heat exchanger and air cooler for the auxiliary cooling system. Temperatures of fluids in the auxiliary cooling system were predicted on a scram event from 30 MW operation at 950 degree sign C of the reactor outlet coolant temperature. Under the predicted helium condition of the auxiliary cooling system, integrity of fuel blocks among the core graphite components was investigated by stress analysis. Evaluation results showed that overcooling to the core graphite components and boiling of water in the auxiliary cooling system should be prevented where open area condition of louvers in the air cooler is the full open

  8. 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

  9. Design of fuel failure detection system for multipurpose reactor GA. Siwabessy

    International Nuclear Information System (INIS)

    Sujalmo Saiful; Kuntoro Iman; Sato, Mitsugu; Isshiki, Masahiko.

    1992-01-01

    A fuel failure detection system (FFDS) has been designed for the Reactor GA. Siwabessy. The FFDS is aimed to detect fuel failure by observing delayed neutron released by fission products such as N-17, I-137, Br-87 and Br-88 in the primary cooling system. The delayed neutrons will be detected by using four neutron detectors, type BF-3, which are located inside a Sampling Tank. The detector location has been determined and the location is associated with the transit time from the reactor core outlet to the Sampling Tank, which is approximately 60 seconds. The neutron detection efficiency was calculated by using a computer code named MORSE. The FFDS has the capability to detect as quickly as possible, even a small failure of a fuel element occurring in the reactor core. Therefore the presence of FFDS in a reactor must be considered, in order to prevent further progress if the fuel failure occurs. (author)

  10. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

  11. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes [1000 and 3000 MW(t)] and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950 0 C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950 0 C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG

  12. Conversion of a reactor to partial power output

    International Nuclear Information System (INIS)

    Iljunin, W.G.; Kusnezow, I.A.; Murogow, W.M.; Schmelew, A.N.

    1975-01-01

    The method, among other things, involves an increase in the rate of secondary fissile material production in a fast breeder reactor if the flow of the working fluid through a turbine is reduced as a function of a given amount of reduction of the electric load. This objective will be served by a circuit and circuit variants, respectively, which include a high temperature cooling circuit with, for instance, a sodium cooled HTR, a low temperature cooling circuit with, for instance, a fast or thermal breeder reactor, a working fluid circuit with the turbine, and a heat consumption circuit. In the scheme suggested for operation in the partial power production mode it is envisaged that, as the electric load of the plant decreases the flow of the working fluid upstream of the turbine is kept constant by means of a control system in the working fluid circuit. Additional control systems are used to reduce the amount of heat transmitted by the breeder reactor to the working fluid. The excess amount of heat is distributed to the load connected. This again reduces the temperatures at the inlet and the outlet of the breeder reactor, thus raising its thermal power output. However, the flow through the breeder reactor remains constant all the time. (DG/RF) [de

  13. Standard Problems for CFD Validation for NGNP - Status Report

    International Nuclear Information System (INIS)

    Johnson, Richard W.; Schultz, Richard R.

    2010-01-01

    The U.S. Department of Energy (DOE) is conducting research and development to support the resurgence of nuclear power in the United States 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 project is called the Next Generation Nuclear Plant (NGNP) Project, which is based on a Generation IV reactor concept called the very high temperature reactor (VHTR). The VHTR will be of the prismatic or pebble bed type; the former is considered herein. The VHTR will use helium as the coolant at temperatures ranging from 250 C to perhaps 1000 C. While computational fluid dynamics (CFD) has not previously been used for the safety analysis of nuclear reactors in the United States, it is being considered 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 operational and accident situations. The ''Standard Problem'' is an experimental data set that represents an important physical phenomenon or phenomena, whose selection is based on a phenomena identification and ranking table (PIRT) for the reactor in question. It will be necessary to build a database that contains a number of standard problems for use to validate CFD and systems analysis codes for the many physical problems that will need to be analyzed. The first two standard problems that have been developed for CFD validation consider flow in the lower plenum of the VHTR and bypass flow in the prismatic core. Both involve scaled models built from quartz and designed to be installed in the INL's matched index of refraction (MIR) test facility. The MIR facility employs mineral oil as the working fluid at a constant temperature. At this temperature, the index of refraction of the mineral oil is the same as that of the quartz. This provides an advantage to the

  14. Validation of commercial business lists as a proxy for licensed alcohol outlets.

    Science.gov (United States)

    Carlos, Heather A; Gabrielli, Joy; Sargent, James D

    2017-05-19

    Studies of retail alcohol outlets are restricted to regions due to lack of U.S. national data. Commercial business lists (BL) offer a possible solution, but no data exists to determine if BLs could serve as an adequate proxy for license data. This paper compares geospatial measures of alcohol outlets derived from a commercial BL with license data for a large US state. We validated BL data as a measure of off-premise alcohol outlet density and proximity compared to license data for 5528 randomly selected California residential addresses. We calculated three proximity measures (Euclidean distance, road network travel time and distance) and two density measures (kernel density estimation and the count within a 2-mile radius) for each dataset. The data was acquired in 2015 and processed and analyzed in 2015 and 2016. Correlations and reliabilities between density (correlation 0.98; Cronbach's α 0.97-0.99) and proximity (correlations 0.77-0.86; α 0.87-0.92) measures were high. For proximity, BL data matched license in 55-57% of addresses, overstated distance in 19%, and understated in 24-26%. BL data can serve as a reliable proxy for licensed alcohol outlets, thus extending the work that can be performed in studies on associations between retail alcohol outlets and drinking outcomes.

  15. Bursting Events in Pressure Flushing with Expanding Bottom Outlet Channel within Dam Reservoir

    Directory of Open Access Journals (Sweden)

    soheila Tofighi

    2017-01-01

    Full Text Available Introduction: Currently, large dams in the world, due to the high amount of sediments in the reservoir, especially around the intake, have operational problems. One of the solutions for this problem is pressure flushing. In this type of flushing, a mixture of water and sediment is removed from bottom outlets form dam reservoir and a funnel shaped crater is created in the vicinity of the outlet opening. In laboratory experiments carried out in this study, pressure flushing with the expansion of bottom outlet within the reservoir and its statistical analysis of bursting events were investigated. The structure of the turbulent flow is not fully understood due to their complexity and random nature. Klein et al. Introduced the turbulence bursting in this kind of flow and Nezo and Nakagora suggested that the events resulting from turbulence bursting has a significant effect of transferring the sediment particles. Materials and Methods: For the purposes of this study, the experiments were conducted with a physical model with 7m length, 1.4m width, and 1.5m height, consisting of three parts namely the inlet of the model, the main reservoir, and settling basin. The main reservoir of the model was 5m long and the sediments were placed within this part of the model. The sediment particles were non-cohesive silica with uniform size and with median diameter (d50 1.15mm and geometrics standard deviation (σg 1.37. Experiments carried out with different discharges and water depths above the bottom outlet in different expansion size of outlet channel in constant sediment level of 20cm above the center of the outlet channel. The model was slowly filled with water until the water surface elevation reached to a desired level. The bottom outlet was manually opened, after a while sedimentwere discharged with the water flow in very high concentrations through the outlet channel (sudden discharge and a funnel shaped crater was formed in front of it. After the run of

  16. Application of analysis technology in nuclear plant

    International Nuclear Information System (INIS)

    Takaoka, Keiko; Miura, Hiromi; Umeda, Kenji

    1996-01-01

    Recently, thanks to the rapid improvement of EWS performance, the authors have been able to carry out design evaluation comparatively, easily, utilizing computational fluid dynamics (CFD) technology. The Nuclear Plant Engineering Department has carried out some analyses in the past several years with the main purpose of evaluating the design of nuclear reactor internals. These studies included ''Thermal Hydraulic Analysis for Top Plenum'' and ''Flow Analysis for Lower Plenum''. It is considered to be a special matter in thermal hydraulic analysis of the top plenum that temperature distribution has been estimated with a relatively small number of meshes by means of an imaginary spray nozzle, and in the flow analysis for the lower plenum that flow distribution has been found to change largely, depending on the reactor internals. One of the ways to confirm the safety of nuclear plants, detailed structural analysis, is required for all possible combinations of transient and load conditions during operation. In particular, it is very important to clarify the thermal stress behavior under operating conditions and to evaluate fatigue analysis in accordance with the Code Requirements. However, it is very complicated and it takes a lot of time. A new system was developed which can operate continuously all of the definitions of the analytical model, the analyzation of pressurized thermal and external stress, and editing reports. In this paper, the authors introduce this system and apply it to a pressurized water reactor

  17. FBR type reactor

    International Nuclear Information System (INIS)

    Nagai, Fumio.

    1979-01-01

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

  18. The “School Foodshed”: schools and fast-food outlets in a London borough

    OpenAIRE

    Caraher, M.; Lloyd, S.; Madelin, T.

    2014-01-01

    Purpose – The purpose of this paper is to explore the location of fast-food outlets around secondary schools and the influence of fast-food availability on the food choices of school children in an inner-London borough. \\ud \\ud Design/methodology/approach – A number of methods including: mapping of outlets relative to schools; sampling food; gathering data on secondary school food policies; observing food behaviour in fast food outlets and focus groups with young people. Findings were fed bac...

  19. Fluid moderator control system reactor internals distribution system

    International Nuclear Information System (INIS)

    Fensterer, H.F.; Klassen, W.E.; Veronesi, L.; Boyle, D.E.; Salton, R.B.

    1987-01-01

    This patent describes a spectral shift pressurized water nuclear reactor employing a low neutron moderating fluid for the spectral shift including a reactor pressure vessel, a core comprising a plurality of fuel assemblies, a core support plate, apparatus comprising means for penetrating the reactor vessel for introducing the moderating fluid into the reactor vessel. Means associated with the core support plate for directly distributing the moderating fluid to and from the fuel assemblies comprises at least one inlet flow channel in the core plate; branch inlet feed lines connect to the inlet flow channel in the core plate; vertical inlet flow lines flow connected to the branch inlet feed lines; each vertical flow line communicates with a fuel assembly; the distribution means further comprise lines serving as return flow lines, each of which is connected to one of the fuel assemblies; branch exit flow lines in the core plate flow connected to the return flow lines of the fuel assembly; and at least one outlet flow channel flow connected to the branch exit flow lines; and a flow port interposed between the penetration means and the distribution means for flow connecting the penetration means with the distribution means

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

    Directory of Open Access Journals (Sweden)

    V. I. Solonin

    2014-01-01

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

  1. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

    Purpose: To ensure the safety for the fuel rod failures by adapting plenum springs to function when small forces such as during transportation of fuel rods is exerted and not to function the resilient force when a relatively great force is exerted. Constitution: Between an upper end plug and a plenum spring in a fuel rod, is disposed an insertion member to the lower portion of which is mounted a pin. This pin is kept upright and causes the plenum spring to function resiliently to the pellets against the loads due to accelerations and mechanical vibrations exerted during transportation of the fuel rods. While on the other hand, if a compression force of a relatively high level is exerted to the plenum spring during reactor operation, the pin of the insertion member is buckled and the insertion member is inserted to the inside of the plenum spring, whereby the pellets are allowed to expand freely and the failures in the fuel elements can be prevented. (Moriyama, K.)

  2. Coolant clean up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tajima, Fumio; Iwami, Hiroshi.

    1981-01-01

    Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

  3. Survey on Cooled-Vessel Designs in High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Kim, Min-Hwan; Lee, Won-Jae

    2006-01-01

    The core outlet temperature of the coolant in the high temperature gas-cooled reactors (HTGR) has been increased to improve the overall efficiency of their electricity generation by using the Brayton cycle or their nuclear hydrogen production by using thermo-chemical processes. The increase of the outlet temperature accompanies an increase of the coolant inlet temperature. A high coolant inlet temperature results in an increase of the reactor pressure vessel (RPV) operation temperature. The conventional steels, proven vessel material in light water reactors, cannot be used as materials for the RPV in the elevated temperatures which necessitate its design to account for the creep effects. Some ferritic or martensitic steels like 2 1/4Cr-1Mo and 9Cr-1Mo-V are very well established creep resistant materials for a temperature range of 400 to 550 C. Although these materials have been used in a chemical plant, there is limited experience with using these materials in nuclear reactors. Even though the 2 1/4Cr-1Mo steel was used to manufacture the RPV for HTR-10 of Japan Atomic Energy Agency(JAEA), a large RPV has not been manufactured by using this material or 9Cr-1Mo-V steel. Due to not only its difficulties in manufacturing but also its high cost, the JAEA determined that they would exclude these materials from the GTHTR design. For the above reasons, KAERI has been considering a cooled-vessel design as an option for the RPV design of a NHDD plant (Nuclear Hydrogen Development and Demonstration). In this study, we surveyed several HTGRs, which adopt the cooled-vessel concept for their RPV design, and discussed their design characteristics. The survey results in design considerations for the NHDD cooled-vessel design

  4. Proximity to Fast-Food Outlets and Supermarkets as Predictors of Fast-Food Dining Frequency.

    Science.gov (United States)

    Athens, Jessica K; Duncan, Dustin T; Elbel, Brian

    2016-08-01

    This study used cross-sectional data to test the independent relationship of proximity to chain fast-food outlets and proximity to full-service supermarkets on the frequency of mealtime dining at fast-food outlets in two major urban areas, using three approaches to define access. Interactions between presence of a supermarket and presence of fast-food outlets as predictors of fast-food dining were also tested. Residential intersections for respondents in point-of-purchase and random-digit-dial telephone surveys of adults in Philadelphia, PA, and Baltimore, MD, were geocoded. The count of fast-food outlets and supermarkets within quarter-mile, half-mile, and 1-mile street network buffers around each respondent's intersection was calculated, as well as distance to the nearest fast-food outlet and supermarket. These variables were regressed on weekly fast-food dining frequency to determine whether proximity to fast food and supermarkets had independent and joint effects on fast-food dining. The effect of access to supermarkets and chain fast-food outlets varied by study population. Among telephone survey respondents, supermarket access was the only significant predictor of fast-food dining frequency. Point-of-purchase respondents were generally unaffected by proximity to either supermarkets or fast-food outlets. However, ≥1 fast-food outlet within a 1-mile buffer was an independent predictor of consuming more fast-food meals among point-of-purchase respondents. At the quarter-mile distance, ≥1 supermarket was predictive of fewer fast-food meals. Supermarket access was associated with less fast-food dining among telephone respondents, whereas access to fast-food outlets were associated with more fast-food visits among survey respondents identified at point-of-purchase. This study adds to the existing literature on geographic determinants of fast-food dining behavior among urban adults in the general population and those who regularly consume fast food. Copyright

  5. JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors. (3) Progress of component design

    International Nuclear Information System (INIS)

    Enuma, Yasuhiro; Kawasaki, Nobuchika; Orita, Junichi; Eto, Masao; Miyagawa, Takayuki

    2015-01-01

    In the frame work of generation IV international forum (GIF), safety design criteria (SDC) and safety design guideline (SDG) for the generation IV sodium-cooled fast reactors have been developing in the circumstance of worldwide deployment of SFRs. JAEA, JAPC, MFBR have been investigating design study for JSFR to satisfy SDC in the feasibility study of SDG for Sodium-cooled Fast Reactor (SFR). In addition to the safety measures, maintainability, reparability and manufacturability are taken into account in the JSFR design study. This paper describes the design of main components. Enlargement of the access route for the inspection devices and addition of the access routes were carried out for the reactor structure. The pump-integrated IHX (pump/IHX) was modified for the primary heat exchanger (PHX), which was installed for the decay heat removal in the IHX at the upper plenum, to be removable for improved repair and maintenance. For the steam generator (SG), protective wall tube type design is under investigation as an option with less R and D risks. (author)

  6. Numerical evaluation of flow through a prismatic very high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Barros Filho, Jose A.; Santos, Andre A.C.; Navarro, Moyses A.; Ribeiro, Felipe Lopes

    2011-01-01

    The High-temperature Gas-cooled reactor (HTGR) is a Next Generation Nuclear System that has a good chance to be used as energy generation source in the near future owing to its potential capacity to supply hydrogen without greenhouse gas emission for the future humanity. Recently, improvements in the HTGR design led to the Very High Temperature Reactor (VHTR) concept in which the outlet temperature of the coolant gas reaches to 1000 deg C increasing the efficiency of the hydrogen and electricity generation. Among the core concepts emerging in the VHTR development stands out the prismatic block which uses coated fuel microspheres named TRISO pressed into cylinders and assembled in hexagonal graphite blocks staked to form columns. The graphite blocks contain flow channels around the fuel cylinders for the helium coolant. In this study an analysis is performed using the CFD code CFX 13.0 on a prismatic fuel assembly in order to investigate its thermo-fluid dynamic performance. The simulations were made in a 1/12 fuel element model of the GT-MHR design which was developed by General Atomics. A numerical mesh verification process based on the Grid Convergence Index (GCI) was performed using five progressively refined meshes to assess the numerical uncertainty of the simulation and determine adequate mesh parameters. An analysis was also performed to evaluate different methods to define the inlet and outlet boundary conditions. In this study simulations of models with and without inlet and outlet plena were compared, showing that the presence of the plena offers a more realistic flow distribution. (author)

  7. Endoscopic stenting versus operative gastrojejunostomy for malignant gastric outlet obstruction.

    Science.gov (United States)

    Chandrasegaram, Manju D; Eslick, Guy D; Mansfield, Clare O; Liem, Han; Richardson, Mark; Ahmed, Sulman; Cox, Michael R

    2012-02-01

    Malignant gastric outlet obstruction represents a terminal stage in pancreatic cancer. Between 5% and 25% of patients with pancreatic cancer ultimately experience malignant gastric outlet obstruction. The aim in palliating patients with malignant gastric outlet obstruction is to reestablish an oral intake by restoring gastrointestinal continuity. This ultimately improves their quality of life in the advanced stages of cancer. The main drawback to operative bypass is the high incidence of delayed gastric emptying, particularly in this group of patients with symptomatic obstruction. This study aimed to compare surgical gastrojejunostomy and endoscopic stenting in palliation of malignant gastric outlet obstruction, acknowledging the diversity and heterogeneity of patients with this presentation. This retrospective study investigated patients treated for malignant gastric outlet obstruction from December 1998 to November 2008 at Nepean Hospital, Sydney, Australia. Endoscopic duodenal stenting was performed under fluoroscopic guidance for placement of the stent. The operative patients underwent open surgical gastrojejunostomy. The outcomes assessed included time to diet, hospital length of stay (LOS), biliary drainage procedures, morbidity, and mortality. Of the 45 participants in this study, 26 underwent duodenal stenting and 19 had operative bypass. Comparing the stenting and operative patients, the median time to fluid intake was respectively 0 vs. 7 days (P < 0.001), and the time to intake of solids was 2 vs. 9 days (P = 0.004). The median total LOS was shorter in the stenting group (11 vs. 25 days; P < 0.001), as was the median postprocedure LOS (5 vs. 10 days; P = 0.07). Endoscopic stenting is preferable to operative gastrojejunostomy in terms of shorter LOS, faster return to fluids and solids, and reduced morbidity and in-hospital mortality for patients with a limited life span.

  8. Contamination of faecal coliforms in ice cubes sampled from food outlets in Kubang Kerian, Kelantan.

    Science.gov (United States)

    Noor Izani, N J; Zulaikha, A R; Mohamad Noor, M R; Amri, M A; Mahat, N A

    2012-03-01

    The use of ice cubes in beverages is common among patrons of food outlets in Malaysia although its safety for human consumption remains unclear. Hence, this study was designed to determine the presence of faecal coliforms and several useful water physicochemical parameters viz. free residual chlorine concentration, turbidity and pH in ice cubes from 30 randomly selected food outlets in Kubang Kerian, Kelantan. Faecal coliforms were found in ice cubes in 16 (53%) food outlets ranging between 1 CFU/100mL to >50 CFU/ 100mL, while in the remaining 14 (47%) food outlets, in samples of tap water as well as in commercially bottled drinking water, faecal coliforms were not detected. The highest faecal coliform counts of >50 CFU/100mL were observed in 3 (10%) food outlets followed by 11-50 CFU/100mL and 1-10 CFU/100mL in 7 (23%) and 6 (20%) food outlets, respectively. All samples recorded low free residual chlorine concentration (contamination by faecal coliforms was not detected in 47% of the samples, tap water and commercially bottled drinking water, it was concluded that (1) contamination by faecal coliforms may occur due to improper handling of ice cubes at the food outlets or (2) they may not be the water sources used for making ice cubes. Since low free residual chlorine concentrations were observed (food outlets, including that of ice cube is crucial in ensuring better food and water for human consumption.

  9. An experimental study on the two-phase natural circulation flow through the gap between reactor vessel and insulation under ERVC

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang-Soon; Park, Rae-Joon; Cho, Young-Ro; Kim, Sang-Baik; Kim, Hwan-Yeol; Kim, Hee-dong

    2005-04-01

    As part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of APR1400, T-HERMES-SMALL and HERMES-HALF experiments have been performed. For the T-HERMES-SMALL experiments, an 1/21.6 scaled experimental facility was prepared utilizing the results of a scaling analysis to simulate the APR1400 reactor and insulation system. The liquid mass flow rates driven by natural circulation loop were measured by varying the wall heat flux, upper outlet area and configuration, and water head condition. The experimental data were also compared with numerical ones given by simple loop analysis. And non-heating small-scaled experiments have also been performed to certify the hydraulic similarity of the heating experiments by injecting air equivalent to the steam generated in the heating experimental condition. The HERMES-HALF experiment is a half-scaled / non-heating experimental study on the two-phase natural circulation through the annular gap between the reactor vessel and the insulation. The behaviors of the two-phase natural circulation flow in the insulation gap were observed, and the liquid mass flow rates driven by natural circulation loop were measured by varying the air injection rate, the coolant inlet area and configuration, and the outlet area and also the water head condition of coolant reservoir. From the experimental flow observation, the recirculation flows in the near region of the shear key were identified. At a higher air injection rate condition, higher recirculation flows and choking phenomenon in the near region of the shear key were observed. As the water inlet areas increased, the natural circulation mass flow rates asymptotically increased, that is, they converged at a specific value. And the experimental correlations for the natural circulation mass flow rates along with a variation of the inlet / outlet area and wall heat flux were

  10. Predictable anomalies of process parameters on failure mode of internal structures in RPV by transient thermal-hydraulic analysis

    International Nuclear Information System (INIS)

    Maki, Akira; Mori, Michitsugu; Kanemoto, Shigeru; Konishi, Hideo

    1997-01-01

    A study has been conducted to evaluate how process parameters will exhibit the change in the event of the troubles related to reactor internal by using transient thermal-hydraulic analysis codes (RETRAN3D-MOD002, etc.). In the present study, the following six events are analytically investigated: 1) a leak from the upper plenum; 2) a leak from the middle part of a shroud; 3) a leak from the lower plenum; 4) a leak from the riser pipe for the jet-pump; 5) the blockage of the jet-pump nozzle; and 6) a leak from the jet-pump diffuser. The results by analyses indicated that the leak from the upper plenum resulted in increasing in the inlet temperature of primary loop recirculation (PLR) and in the differential pressure at the core support plate, and decreasing in the neutron flux (reactor power). Similar analyses were made for the five other events to identify the pattern of relevant process parameter variation in each event. (author)

  11. Parallel segmented outlet flow high performance liquid chromatography with multiplexed detection

    International Nuclear Information System (INIS)

    Camenzuli, Michelle; Terry, Jessica M.; Shalliker, R. Andrew; Conlan, Xavier A.; Barnett, Neil W.; Francis, Paul S.

    2013-01-01

    Graphical abstract: -- Highlights: •Multiplexed detection for liquid chromatography. •‘Parallel segmented outlet flow’ distributes inner and outer portions of the analyte zone. •Three detectors were used simultaneously for the determination of opiate alkaloids. -- Abstract: We describe a new approach to multiplex detection for HPLC, exploiting parallel segmented outlet flow – a new column technology that provides pressure-regulated control of eluate flow through multiple outlet channels, which minimises the additional dead volume associated with conventional post-column flow splitting. Using three detectors: one UV-absorbance and two chemiluminescence systems (tris(2,2′-bipyridine)ruthenium(III) and permanganate), we examine the relative responses for six opium poppy (Papaver somniferum) alkaloids under conventional and multiplexed conditions, where approximately 30% of the eluate was distributed to each detector and the remaining solution directed to a collection vessel. The parallel segmented outlet flow mode of operation offers advantages in terms of solvent consumption, waste generation, total analysis time and solute band volume when applying multiple detectors to HPLC, but the manner in which each detection system is influenced by changes in solute concentration and solution flow rates must be carefully considered

  12. Parallel segmented outlet flow high performance liquid chromatography with multiplexed detection

    Energy Technology Data Exchange (ETDEWEB)

    Camenzuli, Michelle [Australian Centre for Research on Separation Science (ACROSS), School of Science and Health, University of Western Sydney (Parramatta), Sydney, NSW (Australia); Terry, Jessica M. [Centre for Chemistry and Biotechnology, School of Life and Environmental Sciences, Deakin University, Geelong, Victoria 3216 (Australia); Shalliker, R. Andrew, E-mail: r.shalliker@uws.edu.au [Australian Centre for Research on Separation Science (ACROSS), School of Science and Health, University of Western Sydney (Parramatta), Sydney, NSW (Australia); Conlan, Xavier A.; Barnett, Neil W. [Centre for Chemistry and Biotechnology, School of Life and Environmental Sciences, Deakin University, Geelong, Victoria 3216 (Australia); Francis, Paul S., E-mail: paul.francis@deakin.edu.au [Centre for Chemistry and Biotechnology, School of Life and Environmental Sciences, Deakin University, Geelong, Victoria 3216 (Australia)

    2013-11-25

    Graphical abstract: -- Highlights: •Multiplexed detection for liquid chromatography. •‘Parallel segmented outlet flow’ distributes inner and outer portions of the analyte zone. •Three detectors were used simultaneously for the determination of opiate alkaloids. -- Abstract: We describe a new approach to multiplex detection for HPLC, exploiting parallel segmented outlet flow – a new column technology that provides pressure-regulated control of eluate flow through multiple outlet channels, which minimises the additional dead volume associated with conventional post-column flow splitting. Using three detectors: one UV-absorbance and two chemiluminescence systems (tris(2,2′-bipyridine)ruthenium(III) and permanganate), we examine the relative responses for six opium poppy (Papaver somniferum) alkaloids under conventional and multiplexed conditions, where approximately 30% of the eluate was distributed to each detector and the remaining solution directed to a collection vessel. The parallel segmented outlet flow mode of operation offers advantages in terms of solvent consumption, waste generation, total analysis time and solute band volume when applying multiple detectors to HPLC, but the manner in which each detection system is influenced by changes in solute concentration and solution flow rates must be carefully considered.

  13. Patterning of neighbourhood food outlets and longitudinal associations with children's eating behaviours.

    Science.gov (United States)

    Timperio, Anna; Crawford, David; Leech, Rebecca M; Lamb, Karen E; Ball, Kylie

    2018-06-01

    This study examined cross-sectional and prospective associations between typologies of neighbourhood food environment and dietary patterns among 10-12 year-old children. Baseline data were collected in 2003 and follow-up data in 2006 from children in Melbourne or Geelong. Parents completed a food frequency questionnaire at both time points. 'Healthful' and 'energy-dense' dietary pattern scores were computed. A Geographic Information System was used to determine the presence or absence of food outlets (cafés/restaurant; fast food; supermarkets/grocery stores; convenience store; greengrocer; and butcher, seafood or poultry retailer) within an 800 m road network buffer of home. Three typologies were identified: 1-variety of food outlets, including those selling core/fresh foods (n = 96); 2-café/restaurant and convenience (n = 160); 3-few types of outlets (n = 208). Latent class analysis was used to identify underlying unobservable typologies of neighbourhood food outlet availability. Linear mixed models were fitted to determine cross-sectional (n = 439) and longitudinal (n = 173) associations between the three identified neighbourhood typologies and each (log-transformed) dietary pattern, accounting for clustering within families and schools. There was little evidence of cross-sectional associations. The longitudinal analyses showed that compared to those with a variety of food outlets, those with few types had 25% lower scores for the healthful dietary pattern (p < 0.05) three years later. For optimal dietary patterns, availability of a variety of food outlets close to home, particularly those where core/fresh foods are available, may be important. Copyright © 2018 Elsevier Inc. All rights reserved.

  14. Tobacco outlet density and tobacco knowledge, beliefs, purchasing behaviours and price among adolescents in Scotland.

    Science.gov (United States)

    Tunstall, Helena; Shortt, Niamh K; Niedzwiedz, Claire L; Richardson, Elizabeth A; Mitchell, Richard J; Pearce, Jamie R

    2018-06-01

    Despite long-term falls in global adult smoking prevalence and over 50 years of tobacco control policies, adolescent smoking persists. Research suggests greater densities of tobacco retail outlets in residential neighbourhoods are associated with higher adolescent smoking rates. Policies to reduce retail outlets have therefore been identified by public health researchers as a potential 'new frontier' in tobacco control. Better understanding of the pathways linking density of tobacco retailers and smoking behaviour could support these policies. In this study we use path analysis to assess how outlet density in the home environment is related to adolescent tobacco knowledge, beliefs, retail purchases and price in Scotland. We assessed 22,049 13 and 15 year old respondents to the nationally representative cross-sectional 2010 Scottish School Adolescent Lifestyle and Substance Use Survey. Outlet density was based on Scottish Tobacco Retailers Register, 2012, data. A spatially-weighted Kernel Density Estimation measure of outlet density within 400 m of respondents' home postcode was grouped into tertiles. The analysis considered whether outlet density was associated with the number of cigarette brands adolescents could name, positive beliefs about smoking, whether smokers purchased cigarettes from shops themselves or through adult proxies and perceived cost of cigarettes. Models were stratified by adolescent smoking status. The path analyses indicated that outlet density was not associated with most outcomes, but small, significant direct effects on knowledge of cigarette brands among those who had never smoked were observed. With each increase in outlet density tertile the mean number of brands adolescents could name rose by 0.07 (mean = 1.60; SD = 1.18; range = 4). This suggests greater outlet densities may have affected adolescents' knowledge of cigarette brands but did not encourage positive attitudes to smoking, purchases from shops or lower cigarette

  15. Calculation of steam content in a draught section of a tank-type boiling water cooled reactor

    International Nuclear Information System (INIS)

    Panajotov, D.P.; Gorburov, V.I.

    1989-01-01

    Structural and hydrodynamic features of a two-phase flow in a draught section of a tank-type boiling water cooled reactor are considered. A calculated model of the steady flow and methods for determining steam content and phase rate profiles under the maximum steam content at the section axis and at some distance from it are proposed. Steam content distribution by height quantitatively agrees with experimental data for the VK-50 reactor. Calculation technique allows one to obtain steam content and phase rate profiles at the section outlet

  16. Remaining Life Estimation Of Secondary Superheater Outlet On Industrial Electrical Boiler

    International Nuclear Information System (INIS)

    Soedardjo; Andryansyah; Arhatari, B.D.; Natsir, Muhammad; Triyadi, Ari; Farokhi

    2001-01-01

    Remaining life estimation of secondary superheater header outlet (SSHO) on industrial electrical boiler has been carried out. Estimation conducted by the observation of microstructure cavitation development based on Neubauer and Wedel theory. The result is available for isolated cavitation development present yet. That Secondary Superheater Outlet component is in good condition after 14 years operated and predicted could be operated for 36 years again

  17. Present status of high-temperature engineering test reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1994-01-01

    The 30MWt HTTR is a high-temperature gas-cooled reactor (HTGR), with a maximum helium coolant temperature of 950degC at the reactor outlet. The construction of the HTTR started in March 1991, with first criticality to be followed in 1998 after commissioning testing. At present the HTTR reactor building (underground part) and its containment vessel have been almost completed and its main components, such as a reactor pressure vessel (RPV), an intermediate heat exchanger, hot gas pipings and graphite core structures, are now manufacturing at their factories at the target of their installation starting in 1994. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. Japan Atomic Energy Research Institute (JAERI) also plans to conduct material and fuel irradiation tests as an innovative basic research after attaining rated power and coolant temperature. Innovative basic researches are now in great request. The paper describes major features of HTTR, present status of its construction and research and test using HTTR. (author)

  18. Present status of High-Temperature engineering Test Reactor (HTTR) program

    International Nuclear Information System (INIS)

    Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

    1993-01-01

    The 30MWt HTTR is a high-temperature gas-cooled reactor (HTGR), with a maximum helium coolant temperature of 950 deg C at the reactor outlet. The construction of the HTTR started in March 1991, with first criticality to be followed in 1998 after commissioning testing. At present the HTTR reactor building (underground part) and its containment vessel have been almost completed and its main components, such as a reactor pressure vessel (RPV), an intermediate heat exchanger, hot gas pipings and graphite core structures, are now manufacturing at their factories at the target of their installation starting in 1994. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. Japan Atomic Energy Research Institute (JAERI) also plans to conduct material and fuel irradiation tests as an innovative basic research after attaining rated power and coolant temperature. Innovative basic researches are now in great request. The paper describes major features of HTTR, present status of its construction and research and test plan using HTTR. (author)

  19. Feedwater recycling system in BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To improve the reactor safety by preventing thermal stresses and cracks generated in structural materials due to the fluctuations in the temperature for high temperature water - low temperature water mixture near the feedwater nozzle. Method: Feedwater pipes are connected to a pressure vessel not directly but by way of a flow control valve. While the recycled water is circulated from an inlet nozzle to an outlet nozzle through a recycle pump, flow control valve and recycling pipeways, feedwater is fed from the feedwater pipes to the recycling pipeways by way of the flow control valve. More specifically, since the high temperature recycle water and the low temperature recycle water are mixed within the pipeways, the temperature fluctuations resulted from the temperature difference between the recycle water and the feedwater is reduced to prevent thermal fatigue and generation of cracks thereby securing the reactor safety. (Furukawa, Y.)

  20. 16 CFR Appendix E to Part 436 - Sample Item 20(4) Table-Status of Company-Owned Outlets

    Science.gov (United States)

    2010-01-01

    ... RULES DISCLOSURE REQUIREMENTS AND PROHIBITIONS CONCERNING FRANCHISING Pt. 436, App. E Appendix E to Part... 5Outlets Reacquired From Franchisees Column 6Outlets Closed Column 7Outlets Sold to Franchisees Column...