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Sample records for two-phase steaming condition

  1. Two-Phase Instability Characteristics of Printed Circuit Steam Generator for the Low Pressure Condition

    International Nuclear Information System (INIS)

    Kang, Han-Ok; Han, Hun Sik; Kim, Young-In; Kim, Keung Koo

    2015-01-01

    Reduction of installation space for steam generators can lead to much smaller reactor vessel with resultant decrease of overall manufacturing cost for the components. A PCHE(Printed Circuit Heat Exchanger) is one of the compact types of heat exchangers available as an alternative to conventional shell and tube heat exchangers. Its name is derived from the procedure used to manufacture the flat metal plates that form the core of the heat exchanger, which is done by chemical milling. These plates are then stacked and diffusion bonded, converting the plates into a solid metal block containing precisely engineered fluid flow passages. PCSG(Printed Circuit Steam Generator) is a potential candidate to be applied to the integral reactor with its compactness and mechanical robustness. For the introduction of new steam generator, design requirement for the two-phase flow instability should be considered. This paper describes two-phase flow instability characteristics of PCSG for the low pressure condition. PCSG is a potential candidate to be applied to the integral reactor with its compactness and mechanical robustness. Interconnecting flow path was developed to mitigate the two-phase flow instability in the cold side. The flow characteristics of two-phase flow instability at the PCSG is examined experimentally in this study

  2. Experimental study of centrifugal pump performance under steam-water two-phase flow conditions at elevated pressures

    International Nuclear Information System (INIS)

    Chan, A.M.C.; Barreca, S.L.; Hartlen, R.T.

    1991-01-01

    The performance of a centrifugal pump under two-phase flow conditions was studied in a closed loop. System voids of increasing magnitude were attained by draining water from the loop in steps. The operating temperature/pressure were varied from 110 degrees C/0.15 MPa to 260 degrees C/4.7 MPa. Only tests in the first quadrant were conducted. In this paper the head-flow characteristics and pump head degradation data are presented and discussed

  3. Two-phase flow induced vibrations in CANDU steam generators

    International Nuclear Information System (INIS)

    Gidi, A.

    2009-01-01

    The U-Bend region of nuclear steam generators tube bundles have suffered from two-phase cross flow induced vibrations. Tubes in this region have experienced high amplitude vibrations leading to catastrophic failures. Turbulent buffeting and fluid-elastic instability has been identified as the main causes. Previous investigations have focused on flow regime and two-phase flow damping ratio. However, tube bundles in steam generators have vapour generated on the surface of the tubes, which might affect the flow regime, void fraction distribution, turbulent intensity levels and tube-flow interaction, all of which have the potential to change the tube vibration response. A cantilevered tube bundle made of electric cartridges heaters was built and tested in a Freon-11 flow loop at McMaster University. Tubes were arranged in a parallel triangular configuration. The bundle was exposed to two-phase cross flows consisting of different combinations of void from two sources, void generated upstream of the bundle and void generated at the surface of the tubes. Tube tip vibration response was measured optically and void fraction was measured by gamma densitometry technique. It was found that tube vibration amplitude in the transverse direction was reduced by a factor of eight for void fraction generated at the tube surfaces only, when compared to the upstream only void generation case. The main explanation for this effect is a reduction in the correlation length of the turbulent buffeting forcing function. Theoretical calculations of the tube vibration response due to turbulent buffeting under the same experimental conditions predicted a similar reduction in tube amplitude. The void fraction for the fluid-elastic instability threshold in the presence of tube bundle void fraction generation was higher than that for the upstream void fraction generation case. The first explanation of this difference is the level of turbulent buffeting forces the tube bundle was exposed to

  4. Two-phase flow field simulation of horizontal steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Rabiee, Ataollah; Kamalinia, Amir Hossein; Hadad, Kamal [School of Mechanical Engineering, Shiraz University, Shiraz (Iran, Islamic Republic of)

    2017-02-15

    The analysis of steam generators as an interface between primary and secondary circuits in light water nuclear power plants is crucial in terms of safety and design issues. VVER-1000 nuclear power plants use horizontal steam generators which demand a detailed thermal hydraulics investigation in order to predict their behavior during normal and transient operational conditions. Two phase flow field simulation on adjacent tube bundles is important in obtaining logical numerical results. However, the complexity of the tube bundles, due to geometry and arrangement, makes it complicated. Employment of porous media is suggested to simplify numerical modeling. This study presents the use of porous media to simulate the tube bundles within a general-purpose computational fluid dynamics code. Solved governing equations are generalized phase continuity, momentum, and energy equations. Boundary conditions, as one of the main challenges in this numerical analysis, are optimized. The model has been verified and tuned by simple two-dimensional geometry. It is shown that the obtained vapor volume fraction near the cold and hot collectors predict the experimental results more accurately than in previous studies.

  5. Instrumentation for localized measurements in two-phase flow conditions

    International Nuclear Information System (INIS)

    Neff, G.G.; Averill, R.H.; Shurts, S.W.

    1979-01-01

    Three types of instrumentation that have been developed by EG and G Idaho, Inc., and its predecessor, Aerojet Nuclear company, at the Idaho National Engineering Laboratory to investigate two-phase flow phenomenon in a nuclear reactor at the Loss-of-Fluid Test (LOFT) facility are discussed: (a) a combination drag disc-turbine transducer (DTT), (b) a multibeam nuclear hardened gamma densitometer system, and (c) a conductivity sensitive liquid level transducer (LLT). The DTT obtains data on the complex problem of two-phase flow conditions in the LOFT primary coolant system during a loss-os-coolant experiment (LOCE). The discussion of the DTT describes how a turbine, measuring coolant velocity, and a drag disc, measuring coolant momentum flux, can provide valuable mass flow data. The nuclear hardened gamma densitometer is used to obtain density and flow regime information for two-phase flow in the LOFT primary coolant system during a LOCE. The LLT is used to measure water and steam conditions within the LOFT reactor core during a LOCE. The LLT design and the type of data obtained are described

  6. Two Phase Flow Stability in the HTR-10 Steam Generator

    Institute of Scientific and Technical Information of China (English)

    居怀明; 左开芬; 刘志勇; 徐元辉

    2001-01-01

    A 10 MW High Temperature Gas Cooled Reactor (HTR-10) designed bythe Institute of Nuclear Energy Technology (INET) is now being constructed. The steam generator (SG) in the HTR-10 is one of the most important components for reactor safety. The thermal-hydraulic performance of the SG was investigated. A full scale HTR-10 Steam Generator Two Tube Engineering Model Test Facility (SGTM-10) was installed and tested at INET. This paper describes the SGTM-10 thermal hydraulic experimental system in detail. The SGTM-10 simulates the actual thermal and structural parameters of the HTR-10. The SGTM-10 includes three separated loops: the primary helium loop, the secondary water loop, and the tertiary cooling water loop. Two parallel tubes are arranged in the test assembly. The main experimental equipment is shown in the paper. Expermental results are given illustrating the effects of the outlet pressures, the heating power, and the inlet subcooling.

  7. Local two-phase modeling of the water-steam flows occurring in steam generators

    International Nuclear Information System (INIS)

    Denefle, Romain

    2013-01-01

    The present study is related to the need of modeling the two-phase flows occurring in a steam generator (liquid at inlet and vapour at outlet). The choice is made to investigate a hybrid modeling of the flow, considering the gas phase as two separated fields, each one being modeled with different closure laws. In so doing, the small and spherical bubbles are modeled through a dispersed approach within the two-fluid model, and the distorted bubbles are simulated with an interface locating method. The main outcome is about the implementation, the verification and the validation of the model dedicated to the large and distorted bubbles, as well as the coupling of the two approaches for the gas, allowing the presentation of demonstration calculations using the so-called hybrid approach. (author)

  8. A double parameters measurement of steam-water two-phase flow with single orifice

    International Nuclear Information System (INIS)

    Zhong Shuoping; Tong Yunxian; Yu Meiying

    1992-08-01

    A double parameters measurement of steam-water two-phase flow with single orifice is described. An on-line measurement device based on micro-computer has been developed. The measured r.m.s error of steam quality is less than 6.5% and the measured relative r.m.s. error of mass flow rate is less than 9%

  9. Experimental study of single- and two-phase flow fields around PWR steam generator tube support plates

    International Nuclear Information System (INIS)

    Bates, J.M.; Stewart, C.W.

    1979-08-01

    Laser-Doppler anemometry (LDA) was used to measure local mean axial velocities and turbulence intnsities at selected locations within a study model dimensionally protypic of an existing PWR steam generator design. The model tube bundle with support plate was installed in a special flow housing that formed part of an isothermal recirculating water flow loop. Flow conditions for this experiment were intended to simulate only typical single-phase flow velocities and were not an attempt to completely model actual steam generator, boiling, two-phase flow conditions. The measurements were performed in water at approximately 85 0 F with test section average velocities of approximately 0.55 and 1.1 fps. These conditions corresponded to Reynolds numbers of approximately 7,000 and approximately 14,000, respectively. Normalized velocity and turbulence intensity ratios are graphically reported. Additional qualitative, photographic investigations of air-water two-phase flows in a PWR steam generator study model were also performed

  10. Investigation for vertical, two-phase steam-water flow of three turbine models

    International Nuclear Information System (INIS)

    Silverman, S.; Goodrich, L.D.

    1977-01-01

    One of the basic quantities of interest during a loss-of-coolant experiment (LOCE) is the primary system mass flow rate. Presently, there are no transducers commercially available which continuously measure this parameter. Therefore, a transducer was designed at EG and G Idaho, Inc. which combines a drag-disc and turbine into a single unit. The basis for the design was that the drag-disc would measure momentum flux (rhoV 2 ), the turbine would measure velocity and the mass flow rate could then be calculated from the two quantities by assuming a flow profile. For two-phase flow, the outputs are approximately proportional to the desired parameter, but rather large errors can be expected under those assumptions. Preliminary evaluation of the experimental two- and single-phase calibration data has resulted in uncertainty estimates of +-8% of range for the turbine and +-20% of range for the drag-disc. In an effort to reduce the errors, further investigations were made to determine what the drag-disc and turbine really measure. In the present paper, three turbine models for vertical, two-phase, steam/water flow are investigated; the Aya Model, the Rouhani Model, and a volumetric flow model. Theoretical predictions are compared with experimental data for vertical, two-phase steam/water flow. For the purposes of the mass flow calculation, velocity profiles were assumed to be flat for the free-field condition. It is appreciated that this may not be true for all cases investigated, but for an initial inspection, flat profiles were assumed

  11. Preliminary Two-Phase Terry Turbine Nozzle Models for RCIC Off-Design Operation Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, James [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-06-12

    This report presents the effort to extend the single-phase analytical Terry turbine model to cover two-phase off-design conditions. The work includes: (1) adding well-established two-phase choking models – the Isentropic Homogenous Equilibrium Model (IHEM) and Moody’s model, and (2) theoretical development and implementation of a two-phase nozzle expansion model. The two choking models provide bounding cases for the two-phase choking mass flow rate. The new two-phase Terry turbine model uses the choking models to calculate the mass flow rate, the critical pressure at the nozzle throat, and steam quality. In the divergent stage, we only consider the vapor phase with a similar model for the single-phase case by assuming that the liquid phase would slip along the wall with a much slower speed and will not contribute the impulse on the rotor. We also modify the stagnation conditions according to two-phase choking conditions at the throat and the cross-section areas for steam flow at the nozzle throat and at the nozzle exit. The new two-phase Terry turbine model was benchmarked with the same steam nozzle test as for the single-phase model. Better agreement with the experimental data is observed than from the single-phase model. We also repeated the Terry turbine nozzle benchmark work against the Sandia CFD simulation results with the two-phase model for the pure steam inlet nozzle case. The RCIC start-up tests were simulated and compared with the single-phase model. Similar results are obtained. Finally, we designed a new RCIC system test case to simulate the self-regulated Terry turbine behavior observed in Fukushima accidents. In this test, a period inlet condition for the steam quality varying from 1 to 0 is applied. For the high quality inlet period, the RCIC system behaves just like the normal operation condition with a high pump injection flow rate and a nominal steam release rate through the turbine, with the net addition of water to the primary system; for

  12. Investigation on two-phase flow instability in steam generator of integrated nuclear reactor

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    In the pressure range of 3-18MPa,high pressure steam-water two-phase flow density wave instability in vertical upward parallel pipes with inner diameter of 12mm is studied experimentally.The oscillation curves of two-phase flow instability and the effects of several parameters on the oscillation threshold of the system are obtained.Based on the small pertubation linearization method and the stability principles of automatic control system,a mathematical model is developed to predict the characteristics of density wave instability threshold.The predictions of the model are in good agreement with the experimental results.

  13. Study on frictional pressure drop of steam-water two phase flow in optimized four-head internal-ribbed tube

    International Nuclear Information System (INIS)

    Wang Weishu; Zhu Xiaojing; Bi Qincheng; Wu Gang; Yu Shuiqing

    2012-01-01

    The optimized internal-ribbed tube is different from the normal internal-ribbed tube on the frictional pressure drop characteristics. The frictional pressure drop characteristics of steam-water two phase flow in horizontal four-head optimized internal-ribbed were studied under adiabatic condition. According to the experimental and calculation results, the two-phase multiplier is greatly affected by the steam quality and pressure. The two-phase multiplier increases with increasing quality, and decreases with increasing pressure. In the near-critical pressure region, the two-phase multiplier is close to 1. The frictional pressure drop of two phase flow in optimized tube is less than that in the normal tube under the same work condition. The good hydrodynamic condition could be achieved when the optimized internal-ribbed tube is used in the heat transfer equipment because the self-compensating characteristics exist due to the reduction of frictional pressure drop. (authors)

  14. Seabrook simulator model upgrade: Implementation and validation of two-phase, nonequilibrium RCS and steam generator models

    International Nuclear Information System (INIS)

    Kao, S.

    1990-01-01

    A number of deficiencies in the original RCS and steam generator models on the Seabrook simulator were found to give unrealistic results under some off-normal and accident conditions. These deficiencies are attributed to the simplistic assumptions used in the original models, such as the homogeneous, equilibrium equations used in the pressurizer and steam generator models, and the single-phase flow model used in the RCS thermal-hydraulic model. To improve the fidelity of the simulator, efforts have been made to upgrade the RCS and steam generator models to include two-phase, nonequilibrium features. In the new RCS model, the following major assumptions are used to derive the finite difference form of the conservation equations: a donor-cell differencing scheme is adopted to allow flow reversal; a single pressure is used to evaluate properties; a single mass flow rate is assumed in each loop; enthalpy is assumed to vary linearly within each control volume; a homogeneous flow is assumed under two-phase conditions. The pressurizer is divided into a vapor region and a liquid region, each of which is represented by a set of mass and energy conservation equations. Interfacial mass and energy exchange mechanisms (condensation and flashing), thermal interactions between the vessel and fluids, and thermal nonequilibrium between the phases are included in the pressurizer model. The steam generator is divided into the vapor dome, riser, and downcomer regions. The assumptions applied are similar to those of the RCS and pressurizer models. A momentum model is incorporated to calculate the recirculation flow and simulate the downcomer level shrink/swell phenomenon. The new RCS and steam generator models are validated by comparing the simulator calculations against sister plant data and FSAR vendor analysis. The results show the new models give realistic and reliable calculations under off-normal and accident conditions

  15. Numerical analysis of gas-liquid two-phase flow in secondary side of steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Murase, Michio; Nakamura, Akira; Yagi, Yoshinori [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2002-09-01

    The steam generator (SG) in a pressurized water reactor (PWR) is an important two-phase flow component as the boundary between the primary loop and the secondary loop. In this study, we performed gas-liquid two-phase flow analyses for SG reliability tests conduced by Nuclear Power Engineering Corporation (NUPEC) using the two-fluid model of a thermal-hydraulic computer code PHOENICS. In order to calculate the location of the boiling initiation accurately, detailed inputs were required for the friction coefficients affecting the velocity distribution and the heat transfer distribution. However, the velocity and heat transfer distributions did not greatly affect the void fractions in the upper region of the heat transfer tubes. The calculated void fractions agreed with the measured values within 4% as the local average and within 2% as an average in a cross-section, except the region of low void fractions. (author)

  16. Experimental research on density wave oscillation of steam-water two-phase flow in parallel inclined internally ribbed pipes

    International Nuclear Information System (INIS)

    Gao Feng; Chen Tingkuan; Luo Yushan; Yin Fei; Liu Weimin

    2005-01-01

    At p=3-10 MPa, G=300-600 kg/(m 2 ·s), Δt sub =30-90 degree C, and q=0-190 kW/m 2 , the experiments on steam-water two-phase flow instabilities have been performed. The test sections are parallel inclined internally ribbed pipes with an outer diameter of φ38.1 mm, a wall thinkness of 7.5 mm, a obliquity of 19.5 and a length more than 15 m length. Based on the experimental results, the effects of pressure, mass velocity, inlet subcooling and asymmetrical heat flux on steam-water two-phase flow density wave oscillation were analyzed. The experimental results showed that the flow system were more stable as pressure increased. As an increase in mass velocity, critical heat flux increased but critical steam quality decreased. Inlet subcooling had a monotone effect on density wave oscillation, when inlet subcooling decreased, critical heat flux decreased. Under a certain working condition, critical heat flux on asymmetrically heating parallel pipes is higher than that on symmetrically heating parallel pipes, that means the system with symmetrically heating parallel pips was more stable. (authors)

  17. R 12 two-phase flow in throttle capillaries in critical flow conditions

    International Nuclear Information System (INIS)

    Petry, G.

    1983-01-01

    In this dissertation, the state of knowledge on two phase flow, its use and measurement processes are given from an extensive search of the literature. In the experimental part of the work, a continuously working experimental circuit was built up, by which single component two phase flow can be examined in critical flow conditions. Using the maintenance equations, a system of equations was produced, by which the content of steam flow, the content of steam volume and the slip between the phases at the end corssection of the capillary can be determined. The transfer of the experimental results into the Baker diagram shows that the experimental values lie in the region of mist, bubble and foam flow. (orig.) [de

  18. Two-phase flow pattern measurements with a wire mesh sensor in a direct steam generating solar thermal collector

    Science.gov (United States)

    Berger, Michael; Mokhtar, Marwan; Zahler, Christian; Willert, Daniel; Neuhäuser, Anton; Schleicher, Eckhard

    2017-06-01

    At Industrial Solar's test facility in Freiburg (Germany), two phase flow patterns have been measured by using a wire mesh sensor from Helmholtz Zentrum Dresden-Rossendorf (HZDR). Main purpose of the measurements was to compare observed two-phase flow patterns with expected flow patterns from models. The two-phase flow pattern is important for the design of direct steam generating solar collectors. Vibrations should be avoided in the peripheral piping, and local dry-outs or large circumferential temperature gradients should be prevented in the absorber tubes. Therefore, the choice of design for operation conditions like mass flow and steam quality are an important step in the engineering process of such a project. Results of a measurement with the wire mesh sensor are the flow pattern and the plug or slug frequency at the given operating conditions. Under the assumption of the collector power, which can be assumed from previous measurements at the same collector and adaption with sun position and incidence angle modifier, also the slip can be evaluated for a wire mesh sensor measurement. Measurements have been performed at different mass flows and pressure levels. Transient behavior has been tested for flashing, change of mass flow, and sudden changes of irradiation (cloud simulation). This paper describes the measurements and the method of evaluation. Results are shown as extruded profiles in top view and in side view. Measurement and model are compared. The tests have been performed at low steam quality, because of the limits of the test facility. Conclusions and implications for possible future measurements at larger collectors are also presented in this paper.

  19. A contribution to the study of two-phase steam-water critical flow

    International Nuclear Information System (INIS)

    Reocreux, M.

    1975-06-01

    Conservation equations were derived to describe two phase flow systems and conditions were established in order to satisfy critical flow. The theoretical analysis performed to establish the above condition has demonstrated the important part played by transfer terms. Experimental studies on glass and metal channels showed the importance of the way evaporation was initiated. (R.L.)

  20. Measurement of Two-Phase Flow Characteristics Under Microgravity Conditions

    Science.gov (United States)

    Keshock, E. G.; Lin, C. S.; Edwards, L. G.; Knapp, J.; Harrison, M. E.; Xhang, X.

    1999-01-01

    This paper describes the technical approach and initial results of a test program for studying two-phase annular flow under the simulated microgravity conditions of KC-135 aircraft flights. A helical coil flow channel orientation was utilized in order to circumvent the restrictions normally associated with drop tower or aircraft flight tests with respect to two-phase flow, namely spatial restrictions preventing channel lengths of sufficient size to accurately measure pressure drops. Additionally, the helical coil geometry is of interest in itself, considering that operating in a microgravity environment vastly simplifies the two-phase flows occurring in coiled flow channels under 1-g conditions for virtually any orientation. Pressure drop measurements were made across four stainless steel coil test sections, having a range of inside tube diameters (0.95 to 1.9 cm), coil diameters (25 - 50 cm), and length-to-diameter ratios (380 - 720). High-speed video photographic flow observations were made in the transparent straight sections immediately preceding and following the coil test sections. A transparent coil of tygon tubing of 1.9 cm inside diameter was also used to obtain flow visualization information within the coil itself. Initial test data has been obtained from one set of KC-135 flight tests, along with benchmark ground tests. Preliminary results appear to indicate that accurate pressure drop data is obtainable using a helical coil geometry that may be related to straight channel flow behavior. Also, video photographic results appear to indicate that the observed slug-annular flow regime transitions agree quite reasonably with the Dukler microgravity map.

  1. CFD-based shape optimization of steam turbine blade cascade in transonic two phase flows

    International Nuclear Information System (INIS)

    Noori Rahim Abadi, S.M.A.; Ahmadpour, A.; Abadi, S.M.N.R.; Meyer, J.P.

    2017-01-01

    Highlights: • CFD-based shape optimization of a nozzle and a turbine blade regarding nucleating steam flow is performed. • Nucleation rate and droplet radius are the best suited objective functions for the optimization process. • Maximum 34% reduction in entropy generation rate is reported for turbine cascade. • A maximum 10% reduction in Baumann factor and a maximum 2.1% increase in efficiency is achieved for a turbine cascade. - Abstract: In this study CFD-based shape optimization of a 3D nozzle and a 2D turbine blade cascade is undertaken in the presence of non-equilibrium condensation within the considered flow channels. A two-fluid formulation is used for the simulation of unsteady, turbulent, supersonic and compressible flow of wet steam accounting for relevant phase interaction between nucleated liquid droplets and continuous vapor phase. An in-house CFD code is developed to solve the governing equations of the two phase flow and was validated against available experimental data. Optimization is carried out in respect to various objective functions. It is shown that nucleation rate and maximum droplet radius are the best suited target functions for reducing thermodynamic and aerodynamic losses caused by the spontaneous nucleation. The maximum increase of 2.1% in turbine blade efficiency is achieved through shape optimization process.

  2. Steam content of the two-phase flow in the Vk-50 boiling water cooled reactor draught section

    International Nuclear Information System (INIS)

    Fedulin, V.N.; Shmelev, V.E.; Solodkij, V.A.; Bartolomej, G.G.

    1983-01-01

    Results are presented of experimental investigation of the two-phase steam-water coolant flow hydrodynamics within the VK-50 reactor draught section. On the basis of the analysis of the obtained data a two-phase coolant flow model in a large diameter channel is proposed. It is shown that the steam-content distribution in the volume of the draught section has a pronounced non-equilibrium character manifested in the steam migration from the periphery to the central region. A minimum value of the steam content at the periphery is attained at the 0.7-1.0 m height; it is followed by a partial steam content levelling over the section. However the total steam content levelling over the cross section of the draught section does not take place. The steam distribution in the water layer over the draught section (overflow zone) is also nonuniform over the reactor section. The non-uniform steam distribution enchances with reduction nn pressure

  3. An experimental study of two-phase flow instability on two parallel channel with low steam quality

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu shaorong; Bo Jinhai; Yao Meisheng; Han Bing; Zhang Youjie

    1988-01-01

    An experimental result of two-phase flow instability on two parallel channel natural circulation with low steam quality is presented. The comparison of instability in the single channel and that in parallel channel is given. The effect of unequal inlet resistance coefficient and unequal power on the parallel channel instability is described and the behaviour of instability with equal exit steam quality in the two channel is investigated

  4. Simulation of a two phase boiling flow in Poseidon geometry with Astrid steam-water software

    International Nuclear Information System (INIS)

    Larrauri, D.

    1997-01-01

    After different validation test runs in tube an annular geometries, the simulation of a subcooled boiling flow in a rod bundle geometry has been achieved with ASTRID Steam-Water software. The experiment we have simulated is the Poseidon experiment. It is a three heating tube geometry. The thermohydraulic conditions of the simulated flow are closed to the DNB conditions. The simulation results are analysed and compared against the available measurements of liquid and wall temperatures. ASTRID Steam-Water behaviour in such a geometry brings satisfaction. The wall and the liquid temperatures are well predicted in the different parts of the flow. The void fraction reaches 40 % in the vicinity of the heating rods. Besides, the evolution of the different calculated variables shows that a three-dimensional simulation gives capital information for the analyse of the physical phenomena involved in this kind of flow. The good results obtained in Poseidon geometry lead us to think about simulating and analyzing rod bundle flows with ASTRID Steam-Water code. (author)

  5. Two-phase flow steam generator simulations on parallel computers using domain decomposition method

    International Nuclear Information System (INIS)

    Belliard, M.

    2003-01-01

    Within the framework of the Domain Decomposition Method (DDM), we present industrial steady state two-phase flow simulations of PWR Steam Generators (SG) using iteration-by-sub-domain methods: standard and Adaptive Dirichlet/Neumann methods (ADN). The averaged mixture balance equations are solved by a Fractional-Step algorithm, jointly with the Crank-Nicholson scheme and the Finite Element Method. The algorithm works with overlapping or non-overlapping sub-domains and with conforming or nonconforming meshing. Computations are run on PC networks or on massively parallel mainframe computers. A CEA code-linker and the PVM package are used (master-slave context). SG mock-up simulations, involving up to 32 sub-domains, highlight the efficiency (speed-up, scalability) and the robustness of the chosen approach. With the DDM, the computational problem size is easily increased to about 1,000,000 cells and the CPU time is significantly reduced. The difficulties related to industrial use are also discussed. (author)

  6. The role of two-phase coolant in moderating fretting in nuclear steam generators

    International Nuclear Information System (INIS)

    Dyke, J.M.

    2004-01-01

    This paper expands the principal of coolant-cushioning in Nuclear Steam Generators whereby the two-phase coolant, especially the bubble film on the tube surface, moderates the vibration of coolant tubes against their supports. The current paper addresses tube bundle and anti-vibration bars (AVB) geometry issues; examines the tube bundle-coolant-AVB interfaces and examines implications for recirculation flow, AVB design and boiler size. In a T(sat) fluid, the tube surface is uniformly coating with growing bubbles whose momentum is perpendicular to the surface at first, then they are swept away by the bulk flow. The combination of this momentum, the phase change and the water film remaining on the surface, counteract the vibration energy of the tube-AVB system, reducing the likelihood of metal-to-metal contact and consequent fretting. To maximize the benefit of the cushioning effect, the following design inputs are needed: 1) the AVB-tube interface should have sufficient clearance for the T(sat) solution to operate, 2) The AVB should be wide enough to generate the necessary cushioning force, and 3) the AVB should be thin enough to be flexible and absorb some of the transferred vibration energy. Furthermore, fretting and crude deposition at the AVB-tube interface can be reduced or eliminated by reducing the number of AVBs, increasing clearances and making the AVBs limber

  7. Operation of pumps in two-phase steam-water flow. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Grison, P; Lauro, J F [Electricite de France, 78 - Chatou

    1978-01-01

    Determining the two-phase flow (critical or not) through a pump is an esential element for a complete description of loss of coolant accident in a PWR reactor. This article descibes the theoretical and experimental research being done on this subject in France. The model of the pump is first described and its behaviour is examined in different possible cases, particularly that of critical flow. The analysis of the behaviour of the pump is then used to define the experimental conditions for the tests. Two test loops, EVA and EPOPEE, were built. The experimental results are then compared with the theoretical forecasts.

  8. Operation of pumps in two-phase steam-water flow

    International Nuclear Information System (INIS)

    Grison, P.; Lauro, J.F.

    1978-01-01

    Determining the two-phase flow (critical or not) through a pump is an esential element for a complete description of loss of coolant accident in a PWR reactor. This article descibes the theoretical and experimental research being done on this subject in France. The model of the pump is first described and its behaviour is examined in different possible cases, particularly that of critical flow. The analysis of the behaviour of the pump is then used to define the experimental conditions for the tests. Two test loops, EVA and EPOPEE, were built. The experimental results are then compared with the theoretical forecasts [fr

  9. Frictional pressure drop of steam-water two-phase flow in helical coils with small helix diameter of HTR-10

    International Nuclear Information System (INIS)

    Bi Qincheng; Chen Tingkuan; Luo Yushan; Zheng Jianxue

    1996-01-01

    Experiments of steam-water two-phase flow frictional pressure drop through five vertically and horizontally positioned helical coils were carried out in the high pressure steam water test loop of Xi'an Jiaotong University. Two kinds of tube with inner diameters of 10 mm and 12 mm were used to form the coils. The helix diameter was 115 mm with coil pitch 22.5 mm. The experimental conditions were: pressure p = 4-14 MPa, mass velocity G = 400-2000 kg/(m 2 ·s), and inner wall heat flux q = 0-750 kW/m 2 . Theoretical analysis with a semi-empirical correlation was made to predict the two-phase flow fictional pressure drop through these kinds of helical coils

  10. Analysis of the Nonlinear Density Wave Two-Phase Instability in a Steam Generator of 600MWe Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Choi, Seok Ki; Kim, Seong O

    2011-01-01

    A 600 MWe demonstration reactor being developed at KAERI employs a once-through helically coiled steam generator. The helically coiled steam generator is compact and is efficient for heat transfer, however, it may suffer from the two-phase instability. It is well known that the density wave instability is the main source of instability among various types of instabilities in a helically coiled S/G in a LMR. In the present study a simple method for analysis of the density wave two phase instability in a liquid metal reactor S/G is proposed and the method is applied to the analysis of density wave instability in a S/G of 600MWe liquid metal reactor

  11. Slug flow transitions in horizontal gas/liquid two-phase flows. Dependence on channel height and system pressure for air/water and steam/water two-phase flows

    International Nuclear Information System (INIS)

    Nakamura, Hideo

    1996-05-01

    The slug flow transitions and related phenomena for horizontal two-phase flows were studied for a better prediction of two-phase flows that typically appear during the reactor loss-of-coolant accidents (LOCAs). For better representation of the flow conditions experimentally, two large-scaled facility: TPTF for high-pressure steam/water two-phase flows and large duct test facility for air/water two-phase flows, were used. The visual observation of the flow using a video-probe was performed in the TPTF experiments for good understanding of the phenomena. The currently-used models and correlations based mostly on the small-scale low-pressure experiments were reviewed and improved based on these experimental results. The modified Taitel-Dukler model for prediction of transition into slug flow from wavy flow and the modified Steen-Wallis correlation for prediction of onset of liquid entrainment from the interfacial waves were obtained. An empirical correlation for the gas-liquid interfacial friction factor was obtained further for prediction of liquid levels at wavy flow. The region of slug flow regime that is generally under influences of the channel height and system pressure was predicted well when these models and correlations were applied together. (author). 90 refs

  12. Evaluation of CFD Methods for Simulation of Two-Phase Boiling Flow Phenomena in a Helical Coil Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Shaver, Dillon [Argonne National Lab. (ANL), Argonne, IL (United States); Liu, Yang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Vegendla, Prasad [Argonne National Lab. (ANL), Argonne, IL (United States); Tentner, Adrian [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-09-30

    The U.S. Department of Energy, Office of Nuclear Energy charges participants in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program with the development of advanced modeling and simulation capabilities that can be used to address design, performance and safety challenges in the development and deployment of advanced reactor technology. The NEAMS has established a high impact problem (HIP) team to demonstrate the applicability of these tools to identification and mitigation of sources of steam generator flow induced vibration (SGFIV). The SGFIV HIP team is working to evaluate vibration sources in an advanced helical coil steam generator using computational fluid dynamics (CFD) simulations of the turbulent primary coolant flow over the outside of the tubes and CFD simulations of the turbulent multiphase boiling secondary coolant flow inside the tubes integrated with high resolution finite element method assessments of the tubes and their associated structural supports. This report summarizes the demonstration of a methodology for the multiphase boiling flow analysis inside the helical coil steam generator tube. A helical coil steam generator configuration has been defined based on the experiments completed by Polytecnico di Milano in the SIET helical coil steam generator tube facility. Simulations of the defined problem have been completed using the Eulerian-Eulerian multi-fluid modeling capabilities of the commercial CFD code STAR-CCM+. Simulations suggest that the two phases will quickly stratify in the slightly inclined pipe of the helical coil steam generator. These results have been successfully benchmarked against both empirical correlations for pressure drop and simulations using an alternate CFD methodology, the dispersed phase mixture modeling capabilities of the open source CFD code Nek5000.

  13. OTSGI--a program analysing two-phase flow instabilities in helical tubes of once-through steam generator

    International Nuclear Information System (INIS)

    Shi Shaoping; Zhou Fangde; Wang Maohua

    1998-01-01

    The author has studied the two-phases flow instabilities of the helical tubes of once-through steam generator. Using liner-frequency-domain analytical method, the authors have derived out a mathematical model and designed the program. In this model, the authors also have considered the thermal dynamic characteristics of the tube's wall. The program is used to calculate the threshold of the stability and the influences of some factors, such as entrance throttling coefficient, system pressure, entrance supercooling degree, et al. The outcomes are compared with other studies

  14. Numerical Simulation of Non-Equilibrium Two-Phase Wet Steam Flow through an Asymmetric Nozzle

    Directory of Open Access Journals (Sweden)

    Miah Md Ashraful Alam

    2017-11-01

    Full Text Available The present study reported of the numerical investigation of a high-speed wet steam flow through an asymmetric nozzle. The spontaneous non-equilibrium homogeneous condensation of wet steam was numerically modeled based on the classical nucleation theory and droplet growth rate equation combined with the field conservations within the computational fluid dynamics (CFD code of ANSYS Fluent 13.0. The equations describing droplet formations and interphase change were solved sequentially after solving the main flow conservation equations. The calculations were carried out assuming the flow two-dimensional, compressible, turbulent, and viscous. The SST k-ω model was used for modeling the turbulence within an unstructured mesh solver. The validation of numerical model was accomplished, and the results showed a good agreement between the numerical simulation and experimental data. The effect of spontaneous non-equilibrium condensation on the jet and shock structures was revealed, and the condensation shown a great influence on the jet structure.

  15. Applications of Pitot-meter techniques in two-phase, steam/water, flow

    International Nuclear Information System (INIS)

    Kastner, W.; Manzano-Ruiz, J.J.

    1985-01-01

    A simple technique, based on the interpretation of dynamic-pressure readings obtained with local and averaging Pitot-meters (APM) in tow-phase flow, is described and analyzed. The mean dynamic-pressure measurements obtained with an APM allow the calculation of the mass flux of the mixture if the steam quality is known and a combination of two slip-factor correlations is used. The local dynamic-pressure measurements with a multiple Pitot-probe technique provide information on the transition between the most commonly found flow patterns in horizontal piping, i.e. stratified, annular and annular-mist flow

  16. Two-phase flow dynamics in a model steam generator under vertical acceleration oscillation field

    International Nuclear Information System (INIS)

    Ishida, T.; Teshima, N.; Sakurai, S.

    1992-01-01

    The influence of periodically varying acceleration on hydrodynamic response has been studied experimentally using an experimental rig which models a marine reactor subject to vertical motion. The effect on the primary loop is small, but the effect on the secondary loop is large. The variables of the secondary loop, such as circulation flow rate and water level, oscillate with acceleration. The variation of gains in frequency response is analysed. The variations of flow in the secondary loop and in the downcome water level, increase in proportion to the acceleration. The effect of the flow resistance in the secondary loop on the two-phase flow dynamics is clarified. (7 figures) (Author)

  17. Analysis of flow induced valve operation and pressure wave propagation for single and two-phase flow conditions

    International Nuclear Information System (INIS)

    Nagel, H.

    1986-01-01

    The flow induced valve operation is calculated for single and two-phase flow conditions by the fluid dynamic computer code DYVRO and results are compared to experimental data. The analysis show that the operational behaviour of the valves is not only dependent on the condition of the induced flow, but also the pipe flow can cause a feedback as a result of the induced pressure waves. For the calculation of pressure wave propagation in pipes of which the operation of flow induced valves has a considerable influence it is therefore necessary to have a coupled analysis of the pressure wave propagation and the operational behaviour of the valves. The analyses of the fast transient transfer from steam to two-phase flow show a good agreement with experimental data. Hence even these very high loads on pipes resulting from such fluid dynamic transients can be calculated realistically. (orig.)

  18. List of benchmarks for simulation tools of steam-water two-phase flows

    Energy Technology Data Exchange (ETDEWEB)

    Mimouni, S. [Electricite de France (EDF), Div. R and D, 78 - Chatou (France); Serre, G. [CEA Grenoble, Dept. de Thermohydraulique et de Physique, DTP, 38 (France)

    2001-07-01

    A physical-numerical benchmarks matrix was drawn up in the context of the ECUME co-development action. Its purpose is to test the different potentialities required for the numerical methods to be used in the codes of the future which will benefit from advanced physics simulations. This benchmarks matrix is to be used for each numerical method in order to answer the following questions: What is the two-phase flow field that the combination of physics model + numerical scheme can process? What is the accuracy of the scheme for each type of physics situation? What is the numerical efficiency (computing time) of the numerical scheme for each type of physics situation? (author)

  19. List of benchmarks for simulation tools of steam-water two-phase flows

    International Nuclear Information System (INIS)

    Mimouni, S.; Serre, G.

    2001-01-01

    A physical-numerical benchmarks matrix was drawn up in the context of the ECUME co-development action. Its purpose is to test the different potentialities required for the numerical methods to be used in the codes of the future which will benefit from advanced physics simulations. This benchmarks matrix is to be used for each numerical method in order to answer the following questions: What is the two-phase flow field that the combination of physics model + numerical scheme can process? What is the accuracy of the scheme for each type of physics situation? What is the numerical efficiency (computing time) of the numerical scheme for each type of physics situation? (author)

  20. CFD Modeling of Wall Steam Condensation: Two-Phase Flow Approach versus Homogeneous Flow Approach

    International Nuclear Information System (INIS)

    Mimouni, S.; Mechitoua, N.; Foissac, A.; Hassanaly, M.; Ouraou, M.

    2011-01-01

    The present work is focused on the condensation heat transfer that plays a dominant role in many accident scenarios postulated to occur in the containment of nuclear reactors. The study compares a general multiphase approach implemented in NEPTUNE C FD with a homogeneous model, of widespread use for engineering studies, implemented in Code S aturne. The model implemented in NEPTUNE C FD assumes that liquid droplets form along the wall within nucleation sites. Vapor condensation on droplets makes them grow. Once the droplet diameter reaches a critical value, gravitational forces compensate surface tension force and then droplets slide over the wall and form a liquid film. This approach allows taking into account simultaneously the mechanical drift between the droplet and the gas, the heat and mass transfer on droplets in the core of the flow and the condensation/evaporation phenomena on the walls. As concern the homogeneous approach, the motion of the liquid film due to the gravitational forces is neglected, as well as the volume occupied by the liquid. Both condensation models and compressible procedures are validated and compared to experimental data provided by the TOSQAN ISP47 experiment (IRSN Saclay). Computational results compare favorably with experimental data, particularly for the Helium and steam volume fractions.

  1. Investigation of temperature fluctuation phenomena in a stratified steam-water two-phase flow in a simulating pressurizer spray pipe of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Miyoshi, Koji, E-mail: miyoshi.koj@inss.co.jp; Takenaka, Nobuyuki; Ishida, Taisuke; Sugimoto, Katsumi

    2017-05-15

    Highlights: • Thermal hydraulics phenomena were discussed in a spray pipe of pressurizer. • Temperature fluctuation was investigated in a stratified steam-water two-phase. • Remarkable liquid temperature fluctuations were observed in the liquid layer. • The observed temperature fluctuations were caused by the internal gravity wave. • The temperature fluctuations decreased with increasing dissolved oxygen. - Abstract: Temperature fluctuation phenomena in a stratified steam-water two-phase flow in a horizontal rectangular duct, which simulate a pressurizer spray pipe of a pressurized water reactor, were studied experimentally. Vertical distributions of the temperature and the liquid velocity were measured with water of various dissolved oxygen concentrations. Large liquid temperature fluctuations were observed when the water was deaerated well and dissolved oxygen concentration was around 10 ppb. The large temperature fluctuations were not observed when the oxygen concentration was higher. It was shown that the observed temperature fluctuations were caused by the internal gravity wave since the Richardson numbers were larger than 0.25 and the temperature fluctuation frequencies were around the Brunt-Väisälä frequencies in the present experimental conditions. The temperature fluctuations decreased by the non-condensable gas since the non-condensable gas suppressed the condensation and the temperature difference in the liquid layer was small.

  2. Thermodynamic analysis of the two-phase ejector air-conditioning system for buses

    International Nuclear Information System (INIS)

    Ünal, Şaban; Yilmaz, Tuncay

    2015-01-01

    Air-conditioning compressors of the buses are usually operated with the power taken from the engine of the buses. Therefore, an improvement in the air-conditioning system will reduce the fuel consumption of the buses. The improvement in the coefficient of performance (COP) of the air-conditioning system can be provided by using the two-phase ejector as an expansion valve in the air-conditioning system. In this study, the thermodynamic analysis of bus air-conditioning system enhanced with a two-phase ejector and two evaporators is performed. Thermodynamic analysis is made assuming that the mixing process in ejector occurs at constant cross-sectional area and constant pressure. The increase rate in the COP with respect to conventional system is analyzed in terms of the subcooling, condenser and evaporator temperatures. The analysis shows that COP improvement of the system by using the two phase ejector as an expansion device is 15% depending on design parameters of the existing bus air-conditioning system. - Highlights: • Thermodynamic analysis of the two-phase ejector refrigeration system. • Analysis of the COP increase rate of bus air-conditioning system. • Analysis of the entrainment ratio of the two-phase ejector refrigeration system

  3. Frictional pressure drop of high pressure steam-water two-phase flow in internally helical ribbed tubes

    International Nuclear Information System (INIS)

    Tingkuan, C.; Xuanzheng, C.

    1987-01-01

    It is well known that the internally helical ribbed tubes are effective in suppressing the dry-out in boiling tubes at high pressures, so they are widely used as furnace water wall tubes in modern large steam power boilers. Design of the boilers requires the data on frictional pressure drop characteristics of the ribbed tubes, but they are not sufficient now. This paper describes the experimental results on the adiabatic frictional pressure drop in both horizontal ribbed tubes with measured mean inside diameter of 11.69 mm and 35.42 mm at high pressure from 10 to 21 MPa, mass flow rate from 350 to 3800 kg/m/sup 2/s and steam quality from 0 to 1 in our high pressure electrically heated water loop. Simultaneously, both smooth tubes under the same conditions for comparison. Based on the tests the correlation for determining the frictional pressure drop of internally ribbed tubes are proposed

  4. Two-phase performance characteristics of a PWR primary pump under LOCA conditions

    International Nuclear Information System (INIS)

    Grison, P.; Lauro, J.F.

    1977-01-01

    A mathematical model, based on the Euler's theory and a limited flashing, is presented for the flow calculation through a pump working in two-phase conditions, Similarity criteria for representative experimental conditions are studied. The experimental test loop and the first experimental results are described. (author)

  5. Influence of the degree of simplification of the two-phase hydrodynamic model on the simulated behaviour dynamics of a steam generator

    International Nuclear Information System (INIS)

    Dupont, J.F.

    1979-03-01

    The principal simplifications of a mathematical model for the simulation of behaviour dynamics of a two-phase flow with heat exchange are examined, as it appears in a steam generator. The theoretical considerations and numerical solutions permit the evaluation of the validity limits and the influence of these simplifications on the results. (G.T.H.)

  6. Application results of a prototype ultrasonic liquid film sensor to a 7 MPa steam-water two-phase flow experiment

    International Nuclear Information System (INIS)

    Aoyama, Goro; Fujimoto, Kiyoshi; Katono, Kenichi; Nagayoshi, Takuji; Baba, Atsushi; Yasuda, Kenichi

    2016-01-01

    A prototype ultrasonic liquid film sensor was applied to a high-temperature steam-water two-phase flow experiment. The liquid film sensor was vertically installed in a loop which was connected to HUSTLE, a multi-purpose steam source test facility. The hydraulic diameter of the measurement section was 9.4 mm. The output waveforms of the sensor were acquired with a digital oscilloscope. The fluid temperature and system pressure were kept at 288°C and 7.2 MPa, respectively, during the experiment. The pulse-echo method was used to calculate the liquid film thickness. The cross-correlation calculation was utilized to determine the time difference between the pulse reflected at the sensor surface and the pulse reflected at the liquid film surface. The time-averaged liquid film thicknesses were less than 0.055 mm in the annular flow condition. The increase of the time-averaged thickness was small with the change of the gas momentum flux. The film thicknesses measured with the sensor were compared with the past experimental results; the former were smaller than one-fourth of the thickness estimated as the mean film thickness. The comparison results suggested that the continuous liquid sublayer thickness was measured with the liquid film sensor. (author)

  7. An investigation of two-dimensional, two-phase flow of steam in a cascade of turbine blading by the time-marching method

    International Nuclear Information System (INIS)

    Teymourtash, A. R.; Mahpeykar, M. R.

    2003-01-01

    During the course of expansion in turbines, the steam at first super cools and then nucleated to become a two-phase mixture. This is an area where greater understanding can lead to improved design. This paper describes a numerical method for the solution of two-dimensional two-phase flow of steam in a cascade of turbine blading; the unsteady euler equations governing the overall behaviour of the fluid are combined with equations describing droplet behaviour and treated by Jasmine fourth order runge Kutta time marching scheme which modified to allow for two-phase effects. The theoretical surface pressure distributions, droplet radii and contours of constant wetness fraction are presented and results are discussed in the light of knowledge of actual surface pressure distributions

  8. Entrainment and deposition studies in two-phase cross flow: comparison between air-water and steam-water in a square horizontal duct. Technical report (final)

    International Nuclear Information System (INIS)

    Berryman, R.J.; Ralph, J.C.; Wade, C.D.

    1981-03-01

    Air-water simulation studies of two phase steam water flow relevant to the upper plenum of a PWR during reflood situations have recently been undertaken at Harwell for the US Nuclear Regulatory Commission. In order to give confidence that the simulation fluids were capable of modelling the important features of the actual system, a relatively basic comparison experiment has been carried out. Water entrainment and deposition tests have been carried out on a pair of 2.5 cm diameter vertical rods mounted in a cross flow of steam or air in a 10.2 cm x 10.2 cm tunnel. The air and steam systems exhibited similar characteristics to one another. A 'critical' film flowrate was identified for the rods which, once reached, either by injection through the sinters or by entrainment from the main two phase stream, was not exceeded with further water addition. The 'critical' film flowrate decreased with increase of cross flow velocity and was lower for air than steam at the same velocity. The results from the air and steam tests were found to be reasonably well correlated on the basis of the cross flow momentum flux of the air or steam

  9. A simple and rational numerical method of two-phase flow with volume-junction model. 2. The numerical method for general condition of two-phase flow in non-equilibrium states

    International Nuclear Information System (INIS)

    Okazaki, Motoaki

    1997-11-01

    In the previous report, the usefulness of a new numerical method to achieve a rigorous numerical calculation using a simple explicit method with the volume-junction model was presented with the verification calculation for the depressurization of a saturated two-phase mixture. In this report, on the basis of solution method above, a numerical method for general condition of two-phase flow in non-equilibrium states is presented. In general condition of two-phase flow, the combinations of saturated and non-saturated conditions of each phase are considered in the each flow of volume and junction. Numerical evaluation programs are separately prepared for each combination of flow condition. Several numerical calculations of various kinds of non-equilibrium two-phase flow are made to examine the validity of the numerical method. Calculated results showed that the thermodynamic states obtained in different solution schemes were consistent with each other. In the first scheme, the states are determined by using the steam table as a function of pressure and specific enthalpy which are obtained as the solutions of simultaneous equations. In the second scheme, density and specific enthalpy of each phase are directly calculated by using conservation equations of mass and enthalpy of each phase, respectively. Further, no accumulation of error in mass and energy was found. As for the specific enthalpy, two cases of using energy equations for the volume are examined. The first case uses total energy conservation equation and the second case uses the type of the first law of thermodynamics. The results of both cases agreed well. (author)

  10. Study on liquid-metal MHD power generation system with two-phase natural circulation. Applicability to fast reactor conditions

    International Nuclear Information System (INIS)

    Saito, Masaki

    2001-03-01

    Feasibility study of the liquid-metal MHD power generation system combined with the high-density two-phase natural circulation has been performed for the applicability to the simple, autonomic energy conversion system of the liquid-metal cooled fast reactor. The present system has many promising aspects not only in the energy conversion process, but also in safety and economical improvements of the liquid-metal cooled fast reactor. In the previous report, as the first step of the feasibility study, the cycle analyses were performed to examine the effects of the main system parameters on the fundamental characteristics of the system. It was found that the cycle efficiency of the present system is enough competitive with that of the conventional steam turbine system. It was also found that the cycle efficiency depends strongly on the gas-liquid slip ratio in the two-phase flow channel. However, it is very difficult to estimate the gas-liquid slip ratio theoretically, especially in the heavy liquid metal two-phase natural circulation. For example, the effects of MHD load on the two-phase flow characteristics, such as the void fraction and gas-liquid slip ratio are not known well. In the present study, therefore, as the second step of the feasibility study, a series of the experiments were performed to investigate, especially, the effect of MHD load at the single-phase shown-comer flow channel on the characteristics of the two-phase natural circulation. In the first series of the experiments, Woods-metal (Density: 9517 Kg/m 3 ) and nitrogen gas were chosen as the two-phase working fluids. The MHD pressure drop was simulated by the ball valve. The experiments with water and nitrogen gas were also performed to check the effects of the physical properties. From the present experiments, it is found that the average void fraction in the two-phase flow channel is determined by the force balance between the MHD pressure drop, frictional and pressure losses in the tube, and

  11. Friction pressure drop and heat transfer coefficient of two-phase flow in helically coiled tube once-through steam generator for integrated type marine water reactor

    International Nuclear Information System (INIS)

    Nariai, Hideki; Kobayashi, Michiyuki; Matsuoka, Takeshi.

    1982-01-01

    Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report. (author)

  12. Preliminary test of an ultrasonic liquid film sensor for high-temperature steam-water two-phase flow experiments

    International Nuclear Information System (INIS)

    Aoyama, Goro; Nagayoshi, Takuji; Baba, Atsushi

    2014-01-01

    A prototype liquid film sensor for high-temperature steam-water experiments has been developed. The sensor shape simulates a boiling water reactor (BWR) fuel rod. The pulse-echo method can be utilized to measure the thickness of the liquid film covering the sensor surface. A piezoelectric element is soldered onto the inside of the sensor casing which consists of two curved casing pieces. After the piezoelectric element is attached, the two casing pieces are laser welded together. It is confirmed that the temperature rise at the time of the laser welding does not influence soldering of the piezoelectric element. The pressure proof test shows that the sensor can be used at a high-pressure condition of 7 MPa. Simple air-water experiments are done at atmospheric pressure to confirm the liquid film thickness can be measured with the sensor. The fluctuation of the liquid film thickness is satisfactorily captured with the sensor. The minimum and maximum thicknesses are 0.084 and 0.180 mm, respectively. The amplitude of the waveform at 286°C is predicted by the calculation based on the acoustic impedance. It is expected that the sensor is able to measure the liquid film thickness even at BWR operating conditions. (author)

  13. EQUILGAS: Program to estimate temperatures and in situ two-phase conditions in geothermal reservoirs using three combined FT-HSH gas equilibria models

    Science.gov (United States)

    Barragán, Rosa María; Núñez, José; Arellano, Víctor Manuel; Nieva, David

    2016-03-01

    Exploration and exploitation of geothermal resources require the estimation of important physical characteristics of reservoirs including temperatures, pressures and in situ two-phase conditions, in order to evaluate possible uses and/or investigate changes due to exploitation. As at relatively high temperatures (>150 °C) reservoir fluids usually attain chemical equilibrium in contact with hot rocks, different models based on the chemistry of fluids have been developed that allow deep conditions to be estimated. Currently either in water-dominated or steam-dominated reservoirs the chemistry of steam has been useful for working out reservoir conditions. In this context, three methods based on the Fischer-Tropsch (FT) and combined H2S-H2 (HSH) mineral-gas reactions have been developed for estimating temperatures and the quality of the in situ two-phase mixture prevailing in the reservoir. For these methods the mineral buffers considered to be controlling H2S-H2 composition of fluids are as follows. The pyrite-magnetite buffer (FT-HSH1); the pyrite-hematite buffer (FT-HSH2) and the pyrite-pyrrhotite buffer (FT-HSH3). Currently from such models the estimations of both, temperature and steam fraction in the two-phase fluid are obtained graphically by using a blank diagram with a background theoretical solution as reference. Thus large errors are involved since the isotherms are highly nonlinear functions while reservoir steam fractions are taken from a logarithmic scale. In order to facilitate the use of the three FT-HSH methods and minimize visual interpolation errors, the EQUILGAS program that numerically solves the equations of the FT-HSH methods was developed. In this work the FT-HSH methods and the EQUILGAS program are described. Illustrative examples for Mexican fields are also given in order to help the users in deciding which method could be more suitable for every specific data set.

  14. Structural conditions of achieving maximum ductility of two-phase Ni-NiO alloys

    International Nuclear Information System (INIS)

    Grabin, V.V.; Dabizha, E.V.; Movchan, B.A.

    1984-01-01

    A study was made on possibility of increasing ductility of two-phase Ni-NiO alloys, proJuced by traditional technology: ingot smelting, rolling and corresponding annealing for production of grain with certain size. The correlation of mechanical properties of Ni-NiO alloys and pure nickel shows that completion of the structural conJition D--lambda (where D - the average grain diameter, lambda - the value of free path between particles) in two-phase alloys enables: to increase the ultimate strength 1.5 times and preserve the basic level of pure nickel plasticity - at 20 deg C; to increase plasticity 1.4-1.5 times with preserved basic level of pure nickel plasticity - at 800 deg C. The conclusions testify to possibility of controlling mechanical properties of two-phase alloys using structural D and lambda parameters It is proposed that creation of structures with more unifor m particle distribution with respect to sizes will the accompanied by further increase of plasticity under D=lambda condition

  15. Operating conditions of steam generators for LMFBR's

    International Nuclear Information System (INIS)

    Ratzel, W.

    1975-01-01

    Operating conditions considered to be appropriate for a LMFBR steam generator are discussed on the example of the SNR 300. The areas covered are steady state and transient conditions, upset and emergency temperature transients, and requirements due to sodium-water reactions. (author)

  16. Operating conditions of steam generators for LMFBR's

    Energy Technology Data Exchange (ETDEWEB)

    Ratzel, W

    1975-07-01

    Operating conditions considered to be appropriate for a LMFBR steam generator are discussed on the example of the SNR 300. The areas covered are steady state and transient conditions, upset and emergency temperature transients, and requirements due to sodium-water reactions. (author)

  17. A two-fluid two-phase model for thermal-hydraulic analysis of a U-tube steam generator

    International Nuclear Information System (INIS)

    Hung, Huanjen; Chieng, Chingchang; Pei, Baushei; Wang, Songfeng

    1993-01-01

    The Advanced Thermal-Hydraulic Analysis Code for Nuclear Steam Generators (ATHANS) was developed on the basis of the THERMIT-UTSG computer code for U-tube steam generators. The main features of the ATHANS model are as follows: (a) the equations are solved in cylindrical coordinates, (b) the number and the arrangement of the control volumes inside the steam generator can be chosen by the user, (c) the virtual mass effect is incorporated, and (d) the conjugate gradient squared method is employed to accelerate and improve the numerical convergence. The performance of the model is successfully validated by comparison with the test data from a Westinghouse model F steam generator at the Maanshan nuclear power plant. Better agreement with the test data can be obtained by a finer grid system using a cylindrical coordinate system and the virtual mass effect. With these advanced features, ATHANS provides the basic framework for further studies on the problems of steam generators, such as analyses of secondary-side corrosion and tube ruptures

  18. Identification of two-phase flow regimes under variable gravity conditions

    International Nuclear Information System (INIS)

    Kamiel S Gabriel; Huawei Han

    2005-01-01

    Full text of publication follows: Two-phase flow is becoming increasingly important as we move into new and more aggressive technologies in the twenty-first century. Some of its many applications include the design of efficient heat transport systems, the transfer and storage of cryogenic fluids, and condensation and flow boiling processes in heat exchangers and energy transport systems. Two-phase flow has many applications in reduced gravity environments experienced in orbiting spacecraft and earth observation satellites. Examples are heat transport systems, the transfer and storage of cryogenic fluids, and condensation and flow boiling processes in heat exchangers. A concave parallel plate capacitance sensor has been developed to measure void fraction for the purpose of objectively identifying flow regimes. The sensor has been used to collect void-fraction data at microgravity conditions aboard the NASA and ESA zero-gravity aircraft. It is shown that the flow regimes can be objectively determined from the probability density functions of the void fraction signals. It was shown that under microgravity conditions four flow regimes exist: bubbly flow, characterized by discrete gas bubbles flowing in the liquid; slug flow, consisting of Taylor bubbles separated by liquid slugs which may or may not contain several small gas bubbles; transitional flow, characterized by the liquid flowing as a film at the tube wall, and the gas phase flowing in the center with the frequent appearance of chaotic, unstable slugs; and annular flow in which the liquid flows as a film along the tube wall and the gas flows uninterrupted through the center. Since many two-phase flow models are flow regime dependent, a method that can accurately and objectively determine flow regimes is required. (authors)

  19. Identification of two-phase flow regimes under variable gravity conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kamiel S Gabriel [University of Ontario Institute of Technology 2000 Simcoe Street North, Oshawa, ON L1H 7K4 (Canada); Huawei Han [Mechanical Engineering Department, University of Saskatchewan 57 Campus Dr., Saskatoon, Saskatchewan, S7N 5A9 (Canada)

    2005-07-01

    Full text of publication follows: Two-phase flow is becoming increasingly important as we move into new and more aggressive technologies in the twenty-first century. Some of its many applications include the design of efficient heat transport systems, the transfer and storage of cryogenic fluids, and condensation and flow boiling processes in heat exchangers and energy transport systems. Two-phase flow has many applications in reduced gravity environments experienced in orbiting spacecraft and earth observation satellites. Examples are heat transport systems, the transfer and storage of cryogenic fluids, and condensation and flow boiling processes in heat exchangers. A concave parallel plate capacitance sensor has been developed to measure void fraction for the purpose of objectively identifying flow regimes. The sensor has been used to collect void-fraction data at microgravity conditions aboard the NASA and ESA zero-gravity aircraft. It is shown that the flow regimes can be objectively determined from the probability density functions of the void fraction signals. It was shown that under microgravity conditions four flow regimes exist: bubbly flow, characterized by discrete gas bubbles flowing in the liquid; slug flow, consisting of Taylor bubbles separated by liquid slugs which may or may not contain several small gas bubbles; transitional flow, characterized by the liquid flowing as a film at the tube wall, and the gas phase flowing in the center with the frequent appearance of chaotic, unstable slugs; and annular flow in which the liquid flows as a film along the tube wall and the gas flows uninterrupted through the center. Since many two-phase flow models are flow regime dependent, a method that can accurately and objectively determine flow regimes is required. (authors)

  20. Two-phase CFD PTS validation in an extended range of thermohydraulics conditions covered by the COSI experiment

    International Nuclear Information System (INIS)

    Coste, P.; Ortolan, A.

    2014-01-01

    Highlights: • Models for large interfaces in two-phase CFD were developed for PTS. • The COSI experiment is used for NEPTUNE C FD integral validation. • COSI is a PWR cold leg scaled 1/100 for volume. • Fifty runs are calculated, covering a large range of flow configurations. • The CFD predicting capability is analysed using global and local measurements. - Abstract: In the context of the Pressurized Water Reactors (PWR) life duration safety studies, some models were developed to address the Pressurized Thermal Shock (PTS) from the two-phase CFD angle, dealing with interfaces much larger than cells size and with direct contact condensation. Such models were implemented in NEPTUNE C FD, a 3D transient Eulerian two-fluid model. The COSI experiment is used for its integral validation. It represents a cold leg scaled 1/100 for volume and power from a 900 MW PWR under a large range of LOCA PTS conditions. In this study, the CFD is evaluated in the whole range of parameters and flow configurations covered by the experiment. In a first step, a single choice of mesh and CFD models parameters is fixed and justified. In a second step, fifty runs are calculated. The CFD predicting capability is analysed, comparing the liquid temperature and the total condensation rate with the experiment, discussing their dependency on the inlet cold liquid rate, on the liquid level in the cold leg and on the difference between co-current and counter-current runs. It is shown that NEPTUNE C FD 1.0.8 calculates with a fair agreement a large range of flow configurations related to ECCS injection and steam condensation

  1. The effect of spacer grid critical component on pressure drop under both single and two phase flow conditions

    International Nuclear Information System (INIS)

    Han, B.; Yang, B.W.; Zhang, H.; Mao, H.; Zha, Y.

    2016-01-01

    As pressure drop is one of the most critical thermal hydraulic parameters for spacer grids the accurate estimation of it is the key to the design and development of spacer grids. Most of the available correlations for pressure drop do not contain any real geometrical parameters that characterize the grid effect. The main functions for spacer grid are structural support and flow mixing. Once the boundary sublayer near the rod bundle is disturbed, the liquid forms swirls or flow separation that affect pressure drop. However, under two phase flow conditions, due to the existence of steam bubble, the complexity for spacer grid are multiplied and pressure drop calculation becomes much more challenging. The influence of the dimple location, distance of mixing vane to the nearest strip, and the effect of inter-subchannel mixing among neighboring subchannels on pressure drop and downstream flow fields are analyzed in this paper. Based on this study, more detailed space grid geometry parameters are recommended for adding into the correlation when predicting pressure drop.

  2. Frictional resistance of adiabatic two-phase flow in narrow rectangular duct under rolling conditions

    International Nuclear Information System (INIS)

    Xing, Dianchuan; Yan, Changqi; Sun, Licheng; Jin, Guangyuan; Tan, Sichao

    2013-01-01

    Highlights: ► Two-phase flow frictional resistance in narrow duct in rolling is studied. ► Frictional resistance behaviors in rolling are divided into three regions. ► Transient frictional pressure drop fluctuates synchronously with rolling motion. ► Conventional correlations are evaluated against experimental data in rolling motion. ► New correlation for transient frictional resistance in rolling motion is developed. - Abstract: Frictional resistance of air-water two-phase flow in a narrow rectangular duct subjected to rolling motion was investigated experimentally. Time-averaged and transient frictional pressure drop under rolling condition were compared with conventional correlation in laminar flow region (Re l l ⩽ 1400) and turbulent flow region (Re l > 1400) respectively. The result shows that, despite no influence on time-averaged frictional resistance, rolling motion does induce periodical fluctuation of the pressure drop in laminar and transition flow regions. Transient frictional pressure drop fluctuates synchronously with the rolling motion both in laminar and in transition flow region, while it is nearly invariable in turbulent flow region. The fluctuation amplitude of the Relative frictional pressure gradient decreases with the increasing of the superficial velocities. Lee and Lee (2002) correlation and Chisholm (1967) correlation could satisfactorily predict time-averaged frictional pressure drop under rolling conditions, whereas poorly predict the transient frictional pressure drop when it fluctuates periodically. A new correlation with better accuracy for predicting the transient frictional pressure drop in rolling motion is achieved by modifying the Chisholm (1967) correlation on the basis of analyzing the present experimental results with a great number of data points

  3. Hydro-dynamic Solute Transport under Two-Phase Flow Conditions.

    Science.gov (United States)

    Karadimitriou, Nikolaos K; Joekar-Niasar, Vahid; Brizuela, Omar Godinez

    2017-07-26

    There are abundant examples of natural, engineering and industrial applications, in which "solute transport" and "mixing" in porous media occur under multiphase flow conditions. Current state-of-the-art understanding and modelling of such processes are established based on flawed and non-representative models. Moreover, there is no direct experimental result to show the true hydrodynamics of transport and mixing under multiphase flow conditions while the saturation topology is being kept constant for a number of flow rates. With the use of a custom-made microscope, and under well-controlled flow boundary conditions, we visualized directly the transport of a tracer in a Reservoir-on-Chip (RoC) micromodel filled with two immiscible fluids. This study provides novel insights into the saturation-dependency of transport and mixing in porous media. To our knowledge, this is the first reported pore-scale experiment in which the saturation topology, relative permeability, and tortuosity were kept constant and transport was studied under different dynamic conditions in a wide range of saturation. The critical role of two-phase hydrodynamic properties on non-Fickian transport and saturation-dependency of dispersion are discussed, which highlight the major flaws in parametrization of existing models.

  4. Evaluations of two-phase natural circulation flow induced in the reactor vessel annular gap under ERVC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang Soon, E-mail: tomo@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Cheung, Fan-Bill [The Pennsylvania State University, University Park, PA 16802 (United States); Park, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2012-12-15

    Highlights: Black-Right-Pointing-Pointer Two-phase natural circulation flow induced in insulation gap was investigated. Black-Right-Pointing-Pointer Half-scaled non-heating experiments were performed to evaluate flow behavior. Black-Right-Pointing-Pointer The loop-integrated momentum equation was formulated and solved asymptotically. Black-Right-Pointing-Pointer First-order approximate solution was obtained and agreed with experimental data. - Abstract: The process of two-phase natural circulation flow induced in the annular gap between the reactor vessel and the insulation under external reactor vessel cooling conditions was investigated experimentally and analytically in this study. HERMES-HALF experiments were performed to observe and quantify the induced two-phase natural circulation flow in the annular gap. A half-scaled non-heating experimental facility was designed by utilizing the results of a scaling analysis to simulate the APR1400 reactor and its insulation system. The behavior of the boiling-induced two-phase natural circulation flow in the annular gap was observed, and the liquid mass flow rates driven by the natural circulation loop and the void fraction distribution were measured. Direct flow visualization revealed that choking would occur under certain flow conditions in the minimum gap region near the shear keys. Specifically, large recirculation flows were observed in the minimum gap region for large air injection rates and small outlet areas. Under such conditions, the injected air could not pass through the minimum gap region, resulting in the occurrence of choking near the minimum gap with a periodical air back flow being generated. Therefore, a design modification of the minimum gap region needs to be done to facilitate steam venting and to prevent choking from occurring. To complement the HERMES-HALF experimental effort, an analytical study of the dependence of the induced natural circulation mass flow rate on the inlet area and the

  5. Test loop for investigation of two-phase-flow in U-tube steam generators with freon as test medium

    International Nuclear Information System (INIS)

    Urbanek, O.; Horvat, M.

    1980-11-01

    Over the last years two projects concerning the thermohydraulic stability of steam generators have been performed at the Institute for Reactor Safety at the Seibersdorf Research Center. The present report first describes the experimental setup that has been used during the project investigated jointly by FZS and VOEST-ALPINE and summarizes some of the results obtained. The larger part of the report deals with an improved experimental apparatus being used at present for investigations performed in cooperation of the Institute fuer Stroemungslehre und Waermeuebertragung of the TU Wien and the FZS taking into account the results of the previous project. (author)

  6. Measurement of mass flux in high temperature high pressure steam-water two-phase flow using a combination of Pitot tubes and a gamma densitometer

    International Nuclear Information System (INIS)

    Chan, A.M.C.; Bzovey, D.

    1990-01-01

    The design and calibration of a two-phase mass-flux measurement device making use of a Pitot-tube rake and a gamma densitometer are described. Five Pitot tubes and three chordal void-fraction measurements are used. Similar devices have been reported previously. The present device is designed for easy operation and simple data interpretation for both axisymmetric and non-axisymmetric flows under high pressure transient two-phase flow conditions. The device was calibrated using a vertical two-phase flow loop as well as a model-scale pump loop in horizontal orientation. Good agreement between the measured two-phase mass fluxes and the single-phase values was obtained in both cases. (orig.)

  7. Experimental and Theoretical Study of Dryout and Post-Dryout Heat Transfer of Steam-Water Two-Phase Flow in the Annular Channel with Narrow Gap

    International Nuclear Information System (INIS)

    Aye Myint

    2004-10-01

    Two-phase annular flow with heat transfer is prevalent in many processes such as industrial and energy reformation processes. Recently, advances in high performance electronic chips and the miniaturisation of electronic circuits in which high heat flux will be created and other compact systems such as Integrated Nuclear Power Device (INPD), the refrigeration/air conditioning, automobile environment control systems have resulted in a great demand for developing efficient heat transfer techniques to accommodate these high heat fluxes. It has been studied by many researchers because of its successful application in many areas, but its influence factor and mechanism of heat transfer remain somewhat unknown yet. In order to understand the heat transfer and flow mechanism in the narrow annular channel, experimental and theoretical study of dryout and post-dryout heat transfer of steam-water two-phase flow in annular channel with narrow gap (1.0 mm and 1.5 mm) have been carried out. The working fluid is deionized water. The range of experimental pressure is 1.0 ∼ 6.OMPa. In correspondence with two different narrow gaps, two kinds of test sections were designed. The test sections were made of specially processed straight stainless steel tubes with linearity error less than 0.01% to form narrow concentric annuli. It also needs a good sealed performance at high pressure and high temperature. The experiments were carried out to investigate the characteristics and occurring conditions of the dryout point. The former Soviet researcher Kutateladse's correlation, based on round tube, was quoted and modified to apply barrow annuli under low flow conditions. At full conditions of the influencing factors, such as geometry of test section, pressure, mass flux, heat flux etc., an empirical correlation was developed to apply to bilaterally heated annuli and it had a good agreement with the experimental data A new analytical model for the dryout point of critical quality in

  8. Two-phase flow regimes and mechanisms of critical heat flux under subcooled flow boiling conditions

    International Nuclear Information System (INIS)

    Le Corre, Jean-Marie; Yao, Shi-Chune; Amon, Cristina H.

    2010-01-01

    A literature review of critical heat flux (CHF) experimental visualizations under subcooled flow boiling conditions was performed and systematically analyzed. Three major types of CHF flow regimes were identified (bubbly, vapor clot and slug flow regime) and a CHF flow regime map was developed, based on a dimensional analysis of the phenomena and available experimental information. It was found that for similar geometric characteristics and pressure, a Weber number (We)/thermodynamic quality (x) map can be used to predict the CHF flow regime. Based on the experimental observations and the review of the available CHF mechanistic models under subcooled flow boiling conditions, hypothetical CHF mechanisms were selected for each CHF flow regime, all based on a concept of wall dry spot overheating, rewetting prevention and subsequent dry spot spreading. Even though the selected concept has not received much attention (in term or theoretical developments and applications) as compared to other more popular DNB models, its basis have often been cited by experimental investigators and is considered by the authors as the 'most-likely' mechanism based on the literature review and analysis performed in this work. The selected modeling concept has the potential to span the CHF conditions from highly subcooled bubbly flow to early stage of annular flow and has been numerically implemented and validated in bubbly flow and coupled with one- and three-dimensional (CFD) two-phase flow codes, in a companion paper. [Le Corre, J.M., Yao, S.C., Amon, C.H., in this issue. A mechanistic model of critical heat flux under subcooled flow boiling conditions for application to one and three-dimensional computer codes. Nucl. Eng. Des.].

  9. Study on liquid-metal MHD power generation system with two-phase natural circulation. Applicability to fast reactor conditions

    International Nuclear Information System (INIS)

    Saito, Masaki

    2000-03-01

    Feasibility study of the liquid-metal MHD power generation system combined with the high-density two-phase natural circulation has been performed for the applicability to the simple, autonomic energy conversion system of the liquid-metal cooled fast reactor. The present system has many promising aspects not only in the energy conversion process, but also in safety and economical improvements of the liquid-metal cooled fast reactor. For example, the high cycle efficiency can be expected because of the similarity of the present cycle to the Ericsson cycle. Sodium-Water Interaction problem can be excluded by proper combination of the working fluids. As the economical feature, the present system is so simple that the liquid-metal main circular pump, the steam turbine generator, and even the steam generator can be excluded if the thermodynamic working fluid is injected directly into the high temperature liquid metal MHD working fluid. In addition, the present system has the potential to be applied to various heat sources including solar energy because of the high flexibility of the operation temperature. In the present paper, as the first step of the feasibility study, the cycle analyses were performed to examine the effects of the main system parameters on the fundamental characteristics of the system. It is found that the cycle efficiency of the present system is enough competitive with that of the conventional steam turbine system. It is, however, found that the cycle efficiency depends strongly on the gas-liquid slip ratio in the two-phase flow channel. As the conclusions, it is recommended to perform experimental study to obtain the fundamental data, such as the gas-liquid slip ratio in the high-density liquid-metal two-phase natural circulation. (author)

  10. Air-lift pumps characteristics under two-phase flow conditions

    International Nuclear Information System (INIS)

    Kassab, Sadek Z.; Kandil, Hamdy A.; Warda, Hassan A.; Ahmed, Wael H.

    2009-01-01

    Air-lift pumps are finding increasing use where pump reliability and low maintenance are required, where corrosive, abrasive, or radioactive fluids in nuclear applications must be handled and when a compressed air is readily available as a source of a renewable energy for water pumping applications. The objective of the present study is to evaluate the performance of a pump under predetermined operating conditions and to optimize the related parameters. For this purpose, an air-lift pump was designed and tested. Experiments were performed for nine submergence ratios, and three risers of different lengths with different air injection pressures. Moreover, the pump was tested under different two-phase flow patterns. A theoretical model is proposed in this study taking into account the flow patterns at the best efficiency range where the pump is operated. The present results showed that the pump capacity and efficiency are functions of the air mass flow rate, submergence ratio, and riser pipe length. The best efficiency range of the air-lift pumps operation was found to be in the slug and slug-churn flow regimes. The proposed model has been compared with experimental data and the most cited models available. The proposed model is in good agreement with experimental results and found to predict the liquid volumetric flux for different flow patterns including bubbly, slug and churn flow patterns

  11. Technical description of the test section SUCOT to investigate a water/steam two-phase flow

    International Nuclear Information System (INIS)

    Daubner, M.; Janssens-Maenhout, G.; Knebel, J.U.

    2002-02-01

    Within the POOLTHY Project of the European Union (Euratom Fourth Framework Programme contract FJ4J-CT95-0003) an active/passive concept (SUCO-Programme) was investigated which controls the heat removal after a potential core melt-down accident in an evolutionary light water reactor by spreading and stabilising the core melt in the reactor sump and flooding the melt with sump water from above. The experiments were performed in the test facility SUCOT which was designed and erected at the Institute for Nuclear and Energy Technologies (IKET). The report gives an overview on the SUCO-Programme and the scaling analysis, which was applied to design the test facility SUCOT. A detailed technical description of the test facility SUCOT is given, in which the natural circulation driven two-phase flow within the reactor sump and relevant phenomena such as flow boiling, disperse bubbly flow with and without mass transfer, and geysering are investigated. The major components of the test facility, the three-loop system and the instrumentation are described. Finally, a perspective for future application of the gained knowledge is given. (orig.) [de

  12. Emergency makeup of nuclear steam generators in blackout conditions

    International Nuclear Information System (INIS)

    Korolev, A.V.; Derevyanko, O.V.

    2014-01-01

    The paper describes an original solution for using steam energy to organize makeup of NPP steam generators in blackout conditions. The proposed solution combines a disk friction turbine and an axial turbine in a single housing to provide a high overall technical effect enabling the replenishment of nuclear steam generators with steam using the pump turbine drive assembly. The application of the design is analyzed and its efficiency and feasibility are shown

  13. Two-phase flow characteristics in BWRs

    International Nuclear Information System (INIS)

    Katono, Kenichi; Aoyama, Goro; Nagayoshi, Takuji; Yasuda, Kenichi; Nishida, Koji

    2014-01-01

    Reliable prediction of two-phase flow characteristics is important for safety and economy improvements of BWR plants. We have been developing two-phase flow measurement tools and techniques for BWR thermal hydraulic conditions, such as a 3D time-averaged X-ray CT system, an ultrasonic liquid film sensor and a wire-mesh sensor. We applied the developed items in experiments using the multi-purpose steam-water test facility known as HUSTLE, which can simulate two-phase thermal-hydraulic conditions in a BWR reactor pressure vessel, and we constructed a detailed instrumentation database. We validated a 3D two-phase flow simulator using the database and developed the reactor internal two-phase flow analysis system. (author)

  14. Effect on two-phase flow frictional pressure drop characteristic in narrow rectangular channel at fluctuant condition

    International Nuclear Information System (INIS)

    Li Changwei; Cao Xiaxin; Sun Licheng; Jin Guangyuan

    2013-01-01

    Based on the data of two-phase flow in narrow rectangular channel, the influence of the two-phase flow friction characteristic under the different fluctuant states was analyzed. Through analyzing the experimental data, it is shown that the fluctuant amplitude of the friction pressure drop is affected slightly by the fluctuant period in narrow rectangular channel, but the frequency of the friction pressure drop fluctuation is changed. However, the change of fluctuant period is of little effect on the average frictional pressure drop. Comparing the φ l 2 (φ g 2 )-X variation curves at static condition with the ones at fluctuant condition, using the L-M method, it's found that the two phase frictional pressure drop in the narrow rectangular channel under the fluctuant state can be calculated by the φ l 2 (φ g 2 )-X variation curve at static condition. (authors)

  15. Prediction of liquid film dryout in two-phase annular-mist flow in a uniformly heated narrow tube development of analytical method under BWR conditions

    International Nuclear Information System (INIS)

    Utsuno, Hideaki; Kaminaga, Fumito

    1998-01-01

    A method was developed based on the conservation lows to predict critical heat flux (CHF) causing liquid film dryout in two-phase annular-mist flow in a uniformly heated narrow tube under BWR conditions. The applicable range of the method is within the pressure of 3-9 MPa, mass flux of 500-2,000 kg/m 2 ·s, heat flux of 0.33-2.0 MW/m 2 and boiling length-to-tube diameter ratio of 200-800. The two-phase annular-mist flow was modeled with the three-fluid streams with liquid film, entrained droplets and gas flow. Governing equations of the method are mass continuity and energy conservation on the three-fluid streams. Constitutive equations on the mass transfer which consist of the entrainment fraction at equilibrium and the mass transfer coefficient were newly proposed in this study. Confirmation of the present method were performed in comparison with the available film flow measurements and various CHF data from experiments in uniformly heated narrow tubes under high pressure steam-water conditions. In the heat flux range (q'' 2 ) practical for a BWR, agreement of the present method with CHF data was obtained as, (Averaged ratio) ± (Standard deviation) = 0.984 ± 0.077, which was shown to be the same or better agreement than the widely-used CHF correlations. (author)

  16. Experimental investigation on local parameter measurement using optical probes in two-phase flow under rolling condition

    International Nuclear Information System (INIS)

    Tian Daogui; Sun Licheng; Yan Changqi; Liu Guoqiang

    2013-01-01

    In order to get more local interfacial information as well as to further comprehend the intrinsic mechanism of two-phase flow under rolling condition, a method was proposed to measure the local parameters by using optical probes under rolling condition in this paper. An experimental investigation of two-phase flow under rolling condition was conducted using the probe fabricated by the authors. It is verified that the probe method is feasible to measure the local parameters in two'-phase flow under rolling condition. The results show that the interfacial parameters distribution near wall region has a distinct periodicity due to the rolling motion. The averaged deviation of the void fraction measured by the probe from that obtained from measured pressure drop is about 8%. (authors)

  17. Operation of a cascade air conditioning system with two-phase loop

    Science.gov (United States)

    Feng, Yinshan; Wang, Jinliang; Zhao, Futao; Verma, Parmesh; Radcliff, Thomas D.

    2018-05-29

    A method of operating a heat transfer system includes starting operation of a first heat transfer fluid vapor/compression circulation loop including a fluid pumping mechanism, a heat exchanger for rejecting thermal energy from a first heat transfer fluid, and a heat absorption side of an internal heat exchanger. A first conduit in a closed fluid circulation loop circulates the first heat transfer fluid therethrough. Operation of a second two-phase heat transfer fluid circulation loop is started after starting operation of the first heat transfer fluid circulation loop. The second heat transfer fluid circulation loop transfers heat to the first heat transfer fluid circulation loop through the internal heat exchanger and includes a heat rejection side of the internal heat exchanger, a liquid pump, and a heat exchanger evaporator. A second conduit in a closed fluid circulation loop circulates a second heat transfer fluid therethrough.

  18. Experimental study of two-phase flow in a proton exchange membrane fuel cell in short-term microgravity condition

    International Nuclear Information System (INIS)

    Guo, Hang; Liu, Xuan; Zhao, Jian Fu; Ye, Fang; Ma, Chong Fang

    2014-01-01

    Highlights: • Two-phase flow in PEMFC cathode channels is observed in different gravity environments. • The PEMFC shows different operating behavior in normal and microgravity conditions. • Water tends can be removed in microgravity conditions at high water production regime. • Liquid aggregation occurs in microgravity conditions at low water production regime. • Effect of gravity on performance and two-phase flow at two operating regimes is studied. - Abstract: Water management is important for improving the performance and stability of proton exchange membrane fuel cells (PEMFCs) for space applications. An in situ visual observation was conducted on the gas–liquid two-phase flow in the cathode channels of a PEMFC in short-term microgravity condition. The microgravity environment was supplied by a drop tower. A single serpentine flow channel with a depth of 2 mm and a width of 2 mm was applied as the cathode flow field. A membrane electrode assembly comprising of a Nafion 112 membrane sandwiched between gas diffusion layers was used. The anode and cathode were loaded with 1 mg cm −2 platinum. The PEMFC shows a distinct operating behavior in microgravity because of the effect of gravity on the two-phase flow. At a high water production regime, cell performance is enhanced by 4.6% and the accumulated liquid water in the flow channel tends can be removed in microgravity conditions to alleviate flooding. At a low water production regime, cell performance deteriorates by 6.6% and liquid aggregation occurs in the flow channel because of the coalescence of dispersed water droplets in microgravity conditions, thus squeezing the flow channel. The operating behavior of PEMFC in microgravity conditions is different from that in normal gravity conditions. Further studies are needed on PEMFC operating characteristics and liquid management for space applications

  19. Mass transport aspects of polymer electrolyte fuel cells under two-phase flow conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, D.

    2007-03-27

    This work deals with selected aspects of mass transport phenomena in PEFCs and DMFCs. Emphasis is placed on the implications originating from the occurrence of two-phase flow within these devices. Optimality of supply, distribution, and removal of the fuel, the oxidant, and the reaction products is of utmost importance for the stability, efficiency, and durability of the devices. Being a prerequisite for high current densities while maintaining sufficient voltage, mass transport optimization contributes to the development of cost effective as well as compact designs and hence competitive fuel cells. [German] Die Visualisierung und Quantifizierung von Fluessigwasseransammlungen in Polymerelektrolytmembran-Brennstoffzellen konnte mittels Neutronenradiographie erreicht werden. Dank dieser neuartigen diagnostischen Methode konnte erstmals die Fluessigwasseransammlung in den poroesen Gasdiffusionsschichten direkt nachgewiesen und quantifiziert werden. Die Kombination von Neutronenradiographie mit ortsaufgeloesten Stromdichtemessungen bzw. lokaler Impedanzspektroskopie erlaubte die Korrelation des inhomogenen Fluessigwasseranfalls mit dem lokalen elektrochemischen Leistungsverhalten. Systematische Untersuchungen an Polymerelektrolyt- und Direkt-Methanol-Brennstoffzellen verdeutlichen sowohl den Einfluss von Betriebsbedingungen als auch die Auswirkung von Materialeigenschaften auf die Ausbildung zweiphasiger Stroemungen.

  20. Two-phase water movement in unsaturated compacted bentonite under isothermal condition

    International Nuclear Information System (INIS)

    Takeuchi, Shinji

    1994-01-01

    Bentonite is considered as one of the most promising buffer materials of engineered barrier system (EBS) for the geological isolation of high level radioactive waste (HLW) in Japan. The EBS may be composed of vitrified waste, overpack and buffer material. In the early stage of setting and backfilling of HLW, a coupled thermal-hydro-mechanical phenomenon may occur in buffer material due to various causes, but water movement may be the most important phenomenon for the coupled process. It is necessary to verify the two-phase movement for the precise modeling of the water movement in unsaturated bentonite. In this study, in order to analyze water movement, the water retention curves and water diffusivity of compacted bentonite were obtained as the functions of water content, dry density and temperature. Also water movement behavior was examined by applying the Philip and de Vries' and Darcy's equations to the obtained water diffusivity. Water potential was measured with a thermocouple psychrometer. The equation for water diffusivity is shown. The measurement of water potential and water diffusivity and the results are reported. (K.I.)

  1. Structural conditions of maximal plasticity of two-phase metal materials

    International Nuclear Information System (INIS)

    Movchan, B.A.

    1975-01-01

    Analysis is given of experimental values of the strength and plasticity of iron- and tungsten-based two-phase materials with the regulated amount of the second phase and the grain size. Specimens in the form of a 120 mm x 200 mm sheet with a thickness of 0.8-1.2 mm are prepared by means of the electron beam evaporation technique and subsequent condensation of the materials on a preheated support. The variable content of the second phase along the sheet in the range 0.5 volume per cent and more than a 10-fold change in the grain size of the metallic matrix are attained by a simultaneous evaporation of pure metal (99.98 per cent) and nonlmetallic material-niobium carbide or zirconium dioxide ZrO 2 -from two separate sources. The content of arbitrarily distributed spherical particles of the second phase corresponding to a maximum of the plasticity depends only on the structural parameter - the d/D ratio. The absolute falue of the plasticity and its dependence on the temperature is a complex function of many variables - mechanical properties of particles and the matrix, peculiarities of interphase interaction on the boundary particle - matrix, the size of particles, the rate of plastic deformation and relaxation processes

  2. Flow-induced vibration of steam generator helical tubes subjected to external liquid cross flow and internal two-phase flow

    International Nuclear Information System (INIS)

    Jong Chull Jo; Myung Jo Jhung; Woong Sik Kim; Hho Jung Kim

    2005-01-01

    Full text of publication follows: This paper addresses the potential flow-induced vibration problems in a helically-coiled tube steam generator of integral-type nuclear reactor, of which the tubes are subjected to liquid cross flow externally and multi-phase flow externally. The thermal-hydraulic conditions of both tube side and shell side flow fields are predicted using a general purpose computational fluid dynamics code employing the finite volume element modeling. To get the natural frequency and corresponding mode shape of the helical type tubes with various conditions, a finite element analysis code is used. Based on the results of both helical coiled tube steam generator thermal-hydraulic and coiled tube modal analyses, turbulence-induced vibration and fluid-elastic instability analyses are performed. And then the potential for damages on the tubes due to either turbulence-induced vibration or fluid-elastic instability is assessed. In the assessment, special emphases are put on the detailed investigation for the effects of support conditions, coil diameter, and helix pitch on the modal, vibration amplitude and instability characteristics of tubes, from which a technical information and basis needed for designers and regulatory reviewers can be derived. (authors)

  3. Review of two-phase instabilities

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Han Ok; Seo, Han Ok; Kang, Hyung Suk; Cho, Bong Hyun; Lee, Doo Jeong

    1997-06-01

    KAERI is carrying out a development of the design for a new type of integral reactors. The once-through helical steam generator is important design features. The study on designs and operating conditions which prevent flow instability should precede the introduction of one-through steam generator. Experiments are currently scheduled to understand two-phase instability, evaluate the effect of each design parameter on the critical point, and determine proper inlet throttling for the prevention of instability. This report covers general two-phase instability with review of existing studies on this topics. The general classification of two phase flow instability and the characteristics of each type of instability are first described. Special attention is paid to BWR core flow instability and once-through steam generator instability. The reactivity feedback and the effect of system parameters are treated mainly for BWR. With relation to once-through steam generators, the characteristics of convective heating and dryout point oscillation are first investigated and then the existing experimental studies are summarized. Finally chapter summarized the proposed correlations for instability boundary conditions. (author). 231 refs., 5 tabs., 47 figs

  4. Understanding the breakdown of classic two-phase theory and spray atomization at engine-relevant conditions

    Science.gov (United States)

    Dahms, Rainer N.

    2016-04-01

    A generalized framework for multi-component liquid injections is presented to understand and predict the breakdown of classic two-phase theory and spray atomization at engine-relevant conditions. The analysis focuses on the thermodynamic structure and the immiscibility state of representative gas-liquid interfaces. The most modern form of Helmholtz energy mixture state equation is utilized which exhibits a unique and physically consistent behavior over the entire two-phase regime of fluid densities. It is combined with generalized models for non-linear gradient theory and for liquid injections to quantify multi-component two-phase interface structures in global thermal equilibrium. Then, the Helmholtz free energy is minimized which determines the interfacial species distribution as a consequence. This minimal free energy state is demonstrated to validate the underlying assumptions of classic two-phase theory and spray atomization. However, under certain engine-relevant conditions for which corroborating experimental data are presented, this requirement for interfacial thermal equilibrium becomes unsustainable. A rigorously derived probability density function quantifies the ability of the interface to develop internal spatial temperature gradients in the presence of significant temperature differences between injected liquid and ambient gas. Then, the interface can no longer be viewed as an isolated system at minimal free energy. Instead, the interfacial dynamics become intimately connected to those of the separated homogeneous phases. Hence, the interface transitions toward a state in local equilibrium whereupon it becomes a dense-fluid mixing layer. A new conceptual view of a transitional liquid injection process emerges from a transition time scale analysis. Close to the nozzle exit, the two-phase interface still remains largely intact and more classic two-phase processes prevail as a consequence. Further downstream, however, the transition to dense-fluid mixing

  5. Understanding the breakdown of classic two-phase theory and spray atomization at engine-relevant conditions

    Energy Technology Data Exchange (ETDEWEB)

    Dahms, Rainer N., E-mail: Rndahms@sandia.gov [Combustion Research Facility, Sandia National Laboratories, Livermore, California 94551 (United States)

    2016-04-15

    A generalized framework for multi-component liquid injections is presented to understand and predict the breakdown of classic two-phase theory and spray atomization at engine-relevant conditions. The analysis focuses on the thermodynamic structure and the immiscibility state of representative gas-liquid interfaces. The most modern form of Helmholtz energy mixture state equation is utilized which exhibits a unique and physically consistent behavior over the entire two-phase regime of fluid densities. It is combined with generalized models for non-linear gradient theory and for liquid injections to quantify multi-component two-phase interface structures in global thermal equilibrium. Then, the Helmholtz free energy is minimized which determines the interfacial species distribution as a consequence. This minimal free energy state is demonstrated to validate the underlying assumptions of classic two-phase theory and spray atomization. However, under certain engine-relevant conditions for which corroborating experimental data are presented, this requirement for interfacial thermal equilibrium becomes unsustainable. A rigorously derived probability density function quantifies the ability of the interface to develop internal spatial temperature gradients in the presence of significant temperature differences between injected liquid and ambient gas. Then, the interface can no longer be viewed as an isolated system at minimal free energy. Instead, the interfacial dynamics become intimately connected to those of the separated homogeneous phases. Hence, the interface transitions toward a state in local equilibrium whereupon it becomes a dense-fluid mixing layer. A new conceptual view of a transitional liquid injection process emerges from a transition time scale analysis. Close to the nozzle exit, the two-phase interface still remains largely intact and more classic two-phase processes prevail as a consequence. Further downstream, however, the transition to dense-fluid mixing

  6. Development of TPNCIRC code for Evaluation of Two-Phase Natural Circulation Flow Performance under External Reactor Vessel Cooling Conditions

    International Nuclear Information System (INIS)

    Choi, A-Reum; Song, Hyuk-Jin; Park, Jong-Woon

    2015-01-01

    During a severe accident, corium is relocated to the lower head of the nuclear reactor pressure vessel (RPV). Design concept of retaining the corium inside a nuclear reactor pressure vessel (RPV) through external cooling under hypothetical core melting accidents is called external reactor vessel cooling (ERVC). In this respect, validated two-phase natural circulation flow (TPNC) model is necessary to determine the adequacy of the ERVC design and operating conditions such as inlet area, form losses, gap distance, riser length and coolant conditions. The most important model generally characterizing the TPNC are void fraction and two-phase friction factors. Typical experimental and analytical studies to be referred to on two-phase circulation flow characteristics are those by Reyes, Gartia et al. based on Vijayan et al., Nayak et al. and Dubey et al. In the present paper, two-phase natural circulation (TPNC) flow characteristics under external reactor vessel cooling (ERVC) conditions are studied using two existing TPNC flow models of Reyes and Gartia et al. incorporating more improved void fraction and two-phase friction models. These models and correlations are integrated into a computer program, TPNCIRC, which can handle candidate ERVC design parameters, such as inlet, riser and downcomer flow lengths and areas, gap size between reactor vessel and surrounding insulations, minor loss factors and operating parameters of decay power, pressure and subcooling. Accuracy of the TPNCIRC program is investigated with respect to the flow rate and void fractions for existing measured data from a general experiment and ULPU specifically designed for the AP1000 in-vessel retention. Also, the effect of some important design parameters are examined for the experimental and plant conditions. Using the flow models and correlations are integrated into a computer program, TPNCIRC, a number of correlations have been examined. This seems coming from the differences of void fractions

  7. Two-phase 1D+1D model of a DMFC: development and validation on extensive operating conditions range

    Energy Technology Data Exchange (ETDEWEB)

    Casalegno, A.; Marchesi, R.; Parenti, D. [Dipartimento di Energetica, Politecnico di Milano (Italy)

    2008-02-15

    A two-phase 1D+1D model of a direct methanol fuel cell (DMFC) is developed, considering overall mass balance, methanol transport in gas phase through anode diffusion layer, methanol and water crossover. The model is quantitatively validated on an extensive range of operating conditions, 24 polarisation curves. The model accurately reproduces DMFC performance in the validation range and, outside this, it is able to predict values under feasible operating conditions. Finally, the estimations of methanol crossover flux are qualitatively and quantitatively similar to experimental measures and the main local quantities' trends are coherent with results obtained with more complex models. (Abstract Copyright [2008], Wiley Periodicals, Inc.)

  8. Influence of the initial conditions for the numerical simulation of two-phase slug flow

    Energy Technology Data Exchange (ETDEWEB)

    Pachas Napa, Alex A.; Morales, Rigoberto E.M.; Medina, Cesar D. Perea

    2010-07-01

    Multiphase flows in pipelines commonly show several patterns depending on the flow rate, geometry and physical properties of the phases. In oil production, the slug flow pattern is the most common among the others. This flow pattern is characterized by an intermittent succession in space and time of an aerated liquid slug and an elongated gas bubble with a liquid film. Slug flow is studied through the slug tracking model described as one-dimensional and Lagrangian frame referenced. In the model, the mass and the momentum balance equations are applied in control volumes constituted by the gas bubble and the liquid slug. Initial conditions must be determined, which need to reproduce the intermittence of the flow pattern. These initial conditions are given by a sequence of flow properties for each unit cell. Properties of the unit cell in initial conditions should reflect the intermittence, for which they can be analyzed in statistical terms. Therefore, statistical distributions should be obtained for the slug flow variables. Distributions are complemented with the mass balance and the bubble design model. The objective of the present work is to obtain initial conditions for the slug tracking model that reproduce a better adjustment of the fluctuating properties for different pipe inclinations (horizontal, vertical or inclined). The numerical results are compared with experimental data obtained by PFG/FEM/UNICAMP for air-water flow at 0 deg, 45 deg and 90 deg and good agreement is observed. (author)

  9. Hybrid-dimensional modelling of two-phase flow through fractured porous media with enhanced matrix fracture transmission conditions

    Science.gov (United States)

    Brenner, Konstantin; Hennicker, Julian; Masson, Roland; Samier, Pierre

    2018-03-01

    In this work, we extend, to two-phase flow, the single-phase Darcy flow model proposed in [26], [12] in which the (d - 1)-dimensional flow in the fractures is coupled with the d-dimensional flow in the matrix. Three types of so called hybrid-dimensional two-phase Darcy flow models are proposed. They all account for fractures acting either as drains or as barriers, since they allow pressure jumps at the matrix-fracture interfaces. The models also permit to treat gravity dominated flow as well as discontinuous capillary pressure at the material interfaces. The three models differ by their transmission conditions at matrix fracture interfaces: while the first model accounts for the nonlinear two-phase Darcy flux conservations, the second and third ones are based on the linear single phase Darcy flux conservations combined with different approximations of the mobilities. We adapt the Vertex Approximate Gradient (VAG) scheme to this problem, in order to account for anisotropy and heterogeneity aspects as well as for applicability on general meshes. Several test cases are presented to compare our hybrid-dimensional models to the generic equi-dimensional model, in which fractures have the same dimension as the matrix, leading to deep insight about the quality of the proposed reduced models.

  10. Investigation of two-phase flow instability under SMART-P core conditions

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun; Lee, Chung Chan

    2005-01-01

    An integral-type advanced light water reactor, named SMART-P, is being continuously studied at KAERI. The reactor core consists of hundreds of closed-channel type fuel assemblies with vertical upward flows. The upper and lower parts of the fuel assembly channels are connected to the common heads. The constant pressure drop imposed on the channel is responsible for the occurrence of density wave oscillations under local boiling and/or natural circulation conditions. The fuel assembly channel with oscillatory flow is highly susceptible to experience the CHF which may cause the fuel failure due to a sudden increase of the cladding temperature. Thus, prevention of the flow instability is an important criterion for the SMART-P core design. Experimental and analytical studies have been conducted in order to investigate the onset of flow instability (OFI) under SMART core conditions. The parallel channel oscillations were observed in a high pressure water-loop test facility. A linear stability analysis model in the frequency-domain was developed for the prediction of the marginal stability boundary (MSB) in the parallel boiling channels

  11. Two-phase flow instability and propagation of disturbances

    International Nuclear Information System (INIS)

    Yadigaroglu, G.

    1984-01-01

    Various mechanisms of static and dynamic macroinstabilities, appearing in two-phase flows, have been considered. Types of instabilities, conditioned by the form of hydraulic characteristics of the channel and density waves are analyzed in detail. Problems of instabilities in nuclear reactor circuits, in particular problems of instabilities, conditioned by water and steam mixing and vapour condensation, and problems of steam generator operation instability are discussed

  12. Two-phase flow models

    International Nuclear Information System (INIS)

    Delaje, Dzh.

    1984-01-01

    General hypothesis used to simplify the equations, describing two-phase flows, are considered. Two-component and one-component models of two-phase flow, as well as Zuber and Findlay model for actual volumetric steam content, and Wallis model, describing the given phase rates, are presented. The conclusion is made, that the two-component model, in which values averaged in time are included, is applicable for the solving of three-dimensional tasks for unsteady two-phase flow. At the same time, using the two-component model, including values, averaged in space only one-dimensional tasks for unsteady two-phase flow can be solved

  13. Fluisd elastic instability and fretting-wear characteristics of steam generator helical tubes subjected to single-phase external flow and two-phase internal flow

    International Nuclear Information System (INIS)

    Jo, Jong Chull; Jhung, Myung Jo; Kim, Woong Sik; Kim, Hho Jung

    2004-01-01

    This study investigates the fluid elastic instability characteristics of steam generator (SG) helical type tubes and the safety assessment of the potential for fretting-wear damages caused by foreign object in operating nuclear power plants. The thermal-hydraulic conditions of both tube side and shell side flow fields are predicted by a general purpose computational fluid dynamics code employing the finite volume element modeling. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for helical type tubes with various conditions. Special emphases are on the effects of coil diameter and the number of turns on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio. Also, the wear rate of helical type tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube. In addition, addressed is the effect of the external pressure on the vibration and fretting wear characteristics of the tube

  14. Non-Darcy interfacial dynamics of air-water two-phase flow in rough fractures under drainage conditions.

    Science.gov (United States)

    Chang, Chun; Ju, Yang; Xie, Heping; Zhou, Quanlin; Gao, Feng

    2017-07-04

    Two-phase flow interfacial dynamics in rough fractures is fundamental to understanding fluid transport in fractured media. The Haines jump of non-Darcy flow in porous media has been investigated at pore scales, but its fundamental processes in rough fractures remain unclear. In this study, the micron-scale Haines jump of the air-water interface in rough fractures was investigated under drainage conditions, with the air-water interface tracked using dyed water and an imaging system. The results indicate that the interfacial velocities represent significant Haines jumps when the meniscus passes from a narrow "throat" to a wide "body", with jump velocities as high as five times the bulk drainage velocity. Locally, each velocity jump corresponds to a fracture aperture variation; statistically, the velocity variations follow an exponential function of the aperture variations at a length scale of ~100 µm to ~100 mm. This spatial-scale-invariant correlation may indicate that the high-speed local velocities during the Haines jump would not average out spatially for a bulk system. The results may help in understanding the origin of interface instabilities and the resulting non-uniform phase distribution, as well as the micron-scale essence of the spatial and temporal instability of two-phase flow in fractured media at the macroscopic scale.

  15. Analytical study of flow instability behaviour in a boiling two-phase natural circulation loop under low quality conditions

    International Nuclear Information System (INIS)

    Nayak, A.K.; Kumar, N.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2002-01-01

    Analytical investigations have been carried out to study the flow instability behaviour in a boiling two-phase natural circulation loop under low quality conditions. For this purpose, the computer code TINFLO-S has been developed. The code solves the conservation equations of mass, momentum and energy and equation of state for homogeneous equilibrium twophase flow using linear analytical technique. The results of the code have been validated with the experimental data of the loop for both the steady state and stability. The study reveals that the stability behaviour of low quality flow oscillations is different from that of the high quality flow oscillations. The instability reduces with increase in power and throttling at the inlet of the heater. The instability first increases and then reduces with increase in pressure at any subcooling. The effects of diameter of riser pipe, heater and the height of the riser on this instability are also investigated. (orig.) [de

  16. New insights on the complex dynamics of two-phase flow in porous media under intermediate-wet conditions.

    Science.gov (United States)

    Rabbani, Harris Sajjad; Joekar-Niasar, Vahid; Pak, Tannaz; Shokri, Nima

    2017-07-04

    Multiphase flow in porous media is important in a number of environmental and industrial applications such as soil remediation, CO 2 sequestration, and enhanced oil recovery. Wetting properties control flow of immiscible fluids in porous media and fluids distribution in the pore space. In contrast to the strong and weak wet conditions, pore-scale physics of immiscible displacement under intermediate-wet conditions is less understood. This study reports the results of a series of two-dimensional high-resolution direct numerical simulations with the aim of understanding the pore-scale dynamics of two-phase immiscible fluid flow under intermediate-wet conditions. Our results show that for intermediate-wet porous media, pore geometry has a strong influence on interface dynamics, leading to co-existence of concave and convex interfaces. Intermediate wettability leads to various interfacial movements which are not identified under imbibition or drainage conditions. These pore-scale events significantly influence macro-scale flow behaviour causing the counter-intuitive decline in recovery of the defending fluid from weak imbibition to intermediate-wet conditions.

  17. Experimental study of gas–liquid two-phase flow through packed bed under natural circulation conditions

    International Nuclear Information System (INIS)

    Chen, Shao-Wen; Miwa, Shuichiro; Griffiths, Matt

    2016-01-01

    Dry-out phenomena in packed beds or porous media may cause a significant digression of cooling/reaction performance in heat transfer/chemical reactor systems. One of the phenomena responsible for the dry-out in packed beds is known as the counter-current flow limitation (CCFL). In order to investigate the CCFL phenomena induced by gas–liquid two-phase flow in packed beds inside a pool, a natural circulation packed bed test facility was designed and constructed. A total of 27 experimental conditions covering various packing media sizes (sphere diameters: 3.0, 6.4 and 9.5 mm), packed bed heights (15, 35 and 50 cm) and water level heights (1.0, 1.5 and 2.0 m) were tested to examine the CCFL criteria with adiabatic air–water two-phase flow under natural circulation conditions. Both CCFL and flow reversal phenomena were observed, and the experimental data including instantaneous and time-averaged void fraction, differential pressure and superficial gas–liquid velocities were collected. The CCFL criteria were determined when periodical oscillations of void fraction and differential pressure appear. In addition, the Wallis correlation for CCFL was utilized for data analysis, and the Wallis coefficient, C, was determined experimentally from the packed bed CCFL tests. Compared to the existing data-sets in literature, the higher C values obtained in the present experiment suggest a possibly higher dry-out heat flux for natural circulation debris systems, which may be due to the water supply from both top and bottom surfaces of the packed beds. Considering the effects of bed height and hydraulic diameter of the packing media, a newly developed model for the Wallis coefficient, C, under natural circulation CCFL is presented. The present model can predict the experimental data with an averaged absolute error of ±7.9%. (author)

  18. Experimental analysis of upward vertical two-phase flow in four-cusp channels simulating the conditions of a typical nuclear reactor channel, degraded by a loss of coolant accident

    International Nuclear Information System (INIS)

    Assad, A.C.A.

    1984-01-01

    The present work deals with an experimental analysis of upward vertical two-phase flow in channels with circular and four-cusp cross-sections. The latter simulates the conditions of a typical nuclear reactor channel, degraded by a loss of coolant accident. Simultaneous flow of air and water has been employed to simulate adiabatic steam-water flow. The installation of air-water separators helped eliminate instabilities during pressure-drop measurements. The gamma ray attenuation was utilized for the void fraction determination. For the four-cusp geommetry, new criteria for two-phase flow regime transitions have been determined, as well as new correlatins for pressure drop and void fraction, as function of the Lockhart-Martinelli factor and vapour mass-fraction, respectively. (Author) [pt

  19. A study on the flow induced vibration in two phase flow under heating and non-heating conditions

    International Nuclear Information System (INIS)

    Kim, Dae Hun

    2007-02-01

    Critical heat flux (CHF) enhancement devices, like a spacer grid with mixing vane, cause flow-induced vibration (FIV) due to turbulence made by structural resistance. CHF enhancement and FIV reduction are usually studied separately. The main purpose of this article is to investigate the relationship between CHF and FIV. Information of flow-induced vibration due to wire coil design, is experimentally presented in this study by detecting flow-induced vibration under the two-phase flow condition with wire coil inserts. CHF experiments were performed in an upward vertical annulus tube under controlled vibration conditions to determine the effect of vibration on CHF. FIV was measured in an upward vertical tube with various wire coil inserts using air-water as flow material. CHF experiments were performed at one atmosphere with mechanically controlled vibration. A quartz tube (inner diameter of 17 mm, thickness of 2mm and length of 0.72 m) was used for outer tube and a SUS-304 tube (outer diameter of 6.35 mm, thickness of 0.89 mm and length of 0.7 m) was used for the inner heater. Vibration of the heater tube with an amplitude range of 0.1 mm to 0.5 mm and a frequency range of 10 Hz to 50 Hz was carried out at a mass flux of 115 kg/m 2 s and 215 kg/m 2 s. CHF was enhanced by vibration with a maximum ratio of 16.4 %. CHF was increased with increased amplitude and quality. The CHF correlation was developed with R (coefficient of correlation) of 0.903. FIV measuring experiments were performed at one atmosphere by changing the inserted wire coil type. An acrylic tube was used for the test section with inner diameter of 25 mm, thickness of 10 mm and length of 0.5 m. Four types of wire coil, which have a thickness of between 2 mm and 3 mm and pitch length of between 25 mm and 50 mm, were used. FIV and dynamic pressure were detected in water mass flux range of 100 ∼ 3060 kg/m 2 s and air mass flux range of 5.02 ∼ 60.3 kg/m 2 s. Vibration increased along with mass flux and

  20. Improvement of Estimation method for two-phase flow in a large-diameter pipe. Pt. 4. Effect of the inlet boundary condition of the upward flow section on flow characteristics

    International Nuclear Information System (INIS)

    Yoneda, Kimitoshi; Okawa, Tomio; Zhou, Shirong

    1999-01-01

    In nuclear power plants, many large-diameter pipes are subject to gas-liquid two-phase flow. For rational design and performance estimation, the flow in the pipes should be predicted accurately. With the correlation used at present, however, the flow analysis can not reach desirable precision. This is partly due to the lack of understanding of the two-phase flow characteristics in large-diameter pipes. Therefore, steam-water two-phase flow in a vertical pipe (155 mm i.d.) was investigated empirically. Lateral distribution data of phase volume fraction, gas velocity and bubble diameter were obtained. The effects of the inlet boundary condition were also observed. The drift velocity in the developing region was considerably affected by the inlet boundary condition. By deriving the correlation of mean bubble diameter as a function of void fraction and pressure, the empirical data was predicted with high accuracy compared with the existing correlation used in best-estimate codes of nuclear reactor safety analysis. (author)

  1. Chaotic oscillations in a low pressure two-phase natural circulation loop under low power and high inlet subcooling conditions

    International Nuclear Information System (INIS)

    Wu, C.Y.; Wang, S.B.; Pan, C.

    1996-01-01

    The oscillation characteristics of a low pressure two-phase natural circulation loop have been investigated experimentally in this study. Experimental results indicate that the characteristics of the thermal hydraulic oscillations can be periodic, with 2-5 fundamental frequencies, or chaotic, depending on the heating power and inlet subcooling. The number of fundamental frequencies of oscillation increases if the inlet subcooling is increased at a given heating power or the heating power is decreased at a given inlet subcooling; chaotic oscillations appear if the inlet subcooling is further increased and/or the heating power is further decreased. A map of the oscillation characteristics is thus established. The change in oscillation characteristics is evident from the time evolution and power spectrum of a thermal hydraulic parameter and the phase portraits of two thermal hydraulic parameters. These results reveal that a strange attractor exists in a low pressure two-phase natural circulation loop with low power and very high inlet subcooling. (orig.)

  2. Failure of fretted steam generator tubes under accident conditions

    International Nuclear Information System (INIS)

    Forrest, C.F.

    1996-10-01

    Tests were carried out with a bank of tubes in a water tunnel to determine the tolerance of flawed nuclear reactor steam generator tubes to accident conditions which would result in high cross-flow velocities. Fourteen specimen tubes were tested, each having one or two types of defect machined into the surface simulating fretting-wear type scars found in some operating steam generators. The tubes were tested at flow velocities sufficient to induce high fluid elastic-type vibrations. Seven of the tubes failed near the thinnest section of the defects during the one-hour tests, due to impacting and/or rubbing between the tube and the support. Strain gauges, displacement transducers, force gauges and an accelerometer were used on the target tube and/or the tube immediately downstream of it to measure their vibrational characteristics

  3. Nonlinear dynamics of two-phase flow

    International Nuclear Information System (INIS)

    Rizwan-uddin

    1986-01-01

    Unstable flow conditions can occur in a wide variety of laboratory and industry equipment that involve two-phase flow. Instabilities in industrial equipment, which include boiling water reactor (BWR) cores, steam generators, heated channels, cryogenic fluid heaters, heat exchangers, etc., are related to their nonlinear dynamics. These instabilities can be of static (Ledinegg instability) or dynamic (density wave oscillations) type. Determination of regions in parameters space where these instabilities can occur and knowledge of system dynamics in or near these regions is essential for the safe operation of such equipment. Many two-phase flow engineering components can be modeled as heated channels. The set of partial differential equations that describes the dynamics of single- and two-phase flow, for the special case of uniform heat flux along the length of the channel, can be reduced to a set of two coupled ordinary differential equations [in inlet velocity v/sub i/(t) and two-phase residence time tau(t)] involving history integrals: a nonlinear ordinary functional differential equation and an integral equation. Hence, to solve these equations, the dependent variables must be specified for -(nu + tau) ≤ t ≤ 0, where nu is the single-phase residence time. This system of nonlinear equations has been solved analytically using asymptotic expansion series for finite but small perturbations and numerically using finite difference techniques

  4. Concept of turbines for ultrasupercritical, supercritical, and subcritical steam conditions

    Science.gov (United States)

    Mikhailov, V. E.; Khomenok, L. A.; Pichugin, I. I.; Kovalev, I. A.; Bozhko, V. V.; Vladimirskii, O. A.; Zaitsev, I. V.; Kachuriner, Yu. Ya.; Nosovitskii, I. A.; Orlik, V. G.

    2017-11-01

    The article describes the design features of condensing turbines for ultrasupercritical initial steam conditions (USSC) and large-capacity cogeneration turbines for super- and subcritical steam conditions having increased steam extractions for district heating purposes. For improving the efficiency and reliability indicators of USSC turbines, it is proposed to use forced cooling of the head high-temperature thermally stressed parts of the high- and intermediate-pressure rotors, reaction-type blades of the high-pressure cylinder (HPC) and at least the first stages of the intermediate-pressure cylinder (IPC), the double-wall HPC casing with narrow flanges of its horizontal joints, a rigid HPC rotor, an extended system of regenerative steam extractions without using extractions from the HPC flow path, and the low-pressure cylinder's inner casing moving in accordance with the IPC thermal expansions. For cogeneration turbines, it is proposed to shift the upper district heating extraction (or its significant part) to the feedwater pump turbine, which will make it possible to improve the turbine plant efficiency and arrange both district heating extractions in the IPC. In addition, in the case of using a disengaging coupling or precision conical bolts in the coupling, this solution will make it possible to disconnect the LPC in shifting the turbine to operate in the cogeneration mode. The article points out the need to intensify turbine development efforts with the use of modern methods for improving their efficiency and reliability involving, in particular, the use of relatively short 3D blades, last stages fitted with longer rotor blades, evaporation techniques for removing moisture in the last-stage diaphragm, and LPC rotor blades with radial grooves on their leading edges.

  5. Technical description of the test section SUCOT to investigate a water/steam two-phase flow; Technische Beschreibung der Testanlage SUCOT zur Untersuchung Wasser/Wasserdampf Zweiphasenstroemung

    Energy Technology Data Exchange (ETDEWEB)

    Daubner, M.; Janssens-Maenhout, G.; Knebel, J.U.

    2002-02-01

    Within the POOLTHY Project of the European Union (Euratom Fourth Framework Programme contract FJ4J-CT95-0003) an active/passive concept (SUCO-Programme) was investigated which controls the heat removal after a potential core melt-down accident in an evolutionary light water reactor by spreading and stabilising the core melt in the reactor sump and flooding the melt with sump water from above. The experiments were performed in the test facility SUCOT which was designed and erected at the Institute for Nuclear and Energy Technologies (IKET). The report gives an overview on the SUCO-Programme and the scaling analysis, which was applied to design the test facility SUCOT. A detailed technical description of the test facility SUCOT is given, in which the natural circulation driven two-phase flow within the reactor sump and relevant phenomena such as flow boiling, disperse bubbly flow with and without mass transfer, and geysering are investigated. The major components of the test facility, the three-loop system and the instrumentation are described. Finally, a perspective for future application of the gained knowledge is given. (orig.) [German] Im Rahmen des POOLTHY Projekts der Europaeischen Union (Euratom Fourth Framework Programme Kennzeichen FJ4J-CT95-0003) wurde ein aktiv/passives Konzept (SUCO-Programm) zur Beherrschung der Nachwaermeabfuhr nach einem potenziellen Kernschmelzunfall in einem evolutionaeren Leichtwasserreaktor untersucht. Das Konzept sieht eine stabile Ausbreitung der Kernschmelze im Reaktorsumpf und deren Kuehlung von oben mit Sumpfwasser vor. Die experimentellen Arbeiten wurden in der dafuer entwickelten und am Institut fuer Kern- und Energietechnik (IKET) gebauten Testanlage SUCOT durchgefuehrt. Der Bericht gibt einen Ueberblick ueber das SUCO-Programm und die Aehnlichkeitsskalierung, die zur Auslegung der Testanlage SUCOT verwendet wurde. Es wird eine detaillierte technische Beschreibung der Testanlage SUCOT gegeben, die zur Untersuchung der durch

  6. Evaluation of steam sterilization conditions for [18F]fludeoxyglucose

    International Nuclear Information System (INIS)

    Santos, Priscilla F.; Nascimento, Leonardo T.; Valente, Eduardo S.; Silva, Juliana B.; Silveira, Marina B.; Ferreira, Soraya Z.

    2011-01-01

    [ 18 F]Flu deoxyglucose ( 18 FDG) is the most commonly used radiopharmaceutical for positron emission tomography. Sterile filtration of the final product into sterile vials using 0.22 μm filter membrane is usually adopted for 18 FDG. However, Good Manufacturing Practice (GMP) guidelines recommend heat sterilization as the method of choice to ensure sterility of pharmaceutical preparations. The aim of this study was to essay different steam sterilization conditions in order to choose the best one for 18 FDG. Three different sterilization conditions were essayed. The first one at 121 deg C for 15 minutes, the second one at 135 deg C for 3.5 minutes and the third one at 133 deg C for 2 minutes. 18 FDG pH-formulation was kept around 6.0. At the end of autoclave cycles, 18 FDG sterility was evaluated by direct inoculation of 18 FDG in culture media and radiochemical purity was determined by TLC and HPLC. Results demonstrated that all essayed conditions were able to ensure 18 FDG sterility, but caused a decrease in radiochemical purity of 18 FDG. Autoclave cycle at 133 deg C for 2 minutes was the best essayed condition for 18 FDG terminal sterilization, once it provided the greater radiochemical purity value and took less time. 18 FDG was able to meet specifications after autoclave cycles, what supports the application of steam sterilization in routine 18 FDG production, in compliance with GMP. (author)

  7. Define optimal conditions for steam generator tube integrity and an extended steam generator service life

    International Nuclear Information System (INIS)

    Lu, Y.C.

    2007-01-01

    Steam generator (SG) tubing materials are susceptible to corrosion degradation in certain electrochemical corrosion potential regions in the presence of some aggressive ions. Because of the hideout of impurities, the local chemistry conditions in areas under sludge and inside SG crevices may be very aggressive with high concentrations of chlorides and other impurities. These areas are the locations where SG tubing materials are susceptible to degradation such as pitting, crevice corrosion, intergranular attack (IGA) and stress corrosion cracking (SCC). The corrosion susceptibility of each SG alloy is different and is a function of the electrochemical corrosion potential (ECP) and chemical environment. Electrochemical corrosion behaviors of major SG tube alloys were studied under some plausible aggressive crevice chemistry conditions. The possible hazardous conditions leading to SG tube degradation and the conditions, which can minimize SG tube degradation have been determined. Optimal operating conditions in the form of a 'Recommended ECP/pH zone' for minimizing corrosion degradation have been defined for all major SG tube materials, including Alloys 600, 800, 690 and 400, under CANDU SG operating and startup conditions. SCC tests and accelerated corrosion tests were carried out to verify and revise the recommended ECP/pH zones. This information is being incorporated into ChemAND, a system health monitor for plant chemistry management developed by AECL, which alloys utilities to evaluate the status of the SG alloys and to minimize SG material degradation by appropriate SG water chemistry management. (author)

  8. Flow visualization of bubble behavior under two-phase natural circulation flow conditions using high speed digital camera

    Energy Technology Data Exchange (ETDEWEB)

    Lemos, Wanderley F.; Su, Jian, E-mail: wlemos@con.ufrj.br, E-mail: sujian@lasme.coppe.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Faccini, Jose L.H., E-mail: faccini@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Lab. de Termo-Hidraulica Experimental

    2013-07-01

    The The present work aims at identifying flow patterns and measuring interfacial parameters in two-phase natural circulation by using visualization technique with high-speed digital camera. The experiments were conducted in the Natural Circulation Circuit (CCN), installed at Nuclear Engineering Institute/CNEN. The thermo-hydraulic circuit comprises heater, heat exchanger, expansion tank, the pressure relief valve and pipes to interconnect the components. A glass tube is installed at the midpoint of the riser connected to the heater outlet. The natural circulation circuit is complemented by acquisition system of values of temperatures, flow and graphic interface. The instrumentation has thermocouples, volumetric flow meter, rotameter and high-speed digital camera. The experimental study is performed through analysis of information from measurements of temperatures at strategic points along the hydraulic circuit, besides natural circulation flow rates. The comparisons between analytical and experimental values are validated by viewing, recording and processing of the images for the flows patterns. Variables involved in the process of identification of flow regimes, dimensionless parameters, the phase velocity of the flow, initial boiling point, the phenomenon of 'flashing' pre-slug flow type were obtained experimentally. (author)

  9. Flow visualization of bubble behavior under two-phase natural circulation flow conditions using high speed digital camera

    International Nuclear Information System (INIS)

    Lemos, Wanderley F.; Su, Jian; Faccini, Jose L.H.

    2013-01-01

    The The present work aims at identifying flow patterns and measuring interfacial parameters in two-phase natural circulation by using visualization technique with high-speed digital camera. The experiments were conducted in the Natural Circulation Circuit (CCN), installed at Nuclear Engineering Institute/CNEN. The thermo-hydraulic circuit comprises heater, heat exchanger, expansion tank, the pressure relief valve and pipes to interconnect the components. A glass tube is installed at the midpoint of the riser connected to the heater outlet. The natural circulation circuit is complemented by acquisition system of values of temperatures, flow and graphic interface. The instrumentation has thermocouples, volumetric flow meter, rotameter and high-speed digital camera. The experimental study is performed through analysis of information from measurements of temperatures at strategic points along the hydraulic circuit, besides natural circulation flow rates. The comparisons between analytical and experimental values are validated by viewing, recording and processing of the images for the flows patterns. Variables involved in the process of identification of flow regimes, dimensionless parameters, the phase velocity of the flow, initial boiling point, the phenomenon of 'flashing' pre-slug flow type were obtained experimentally. (author)

  10. Exact solution for a two-phase Stefan problem with variable latent heat and a convective boundary condition at the fixed face

    Science.gov (United States)

    Bollati, Julieta; Tarzia, Domingo A.

    2018-04-01

    Recently, in Tarzia (Thermal Sci 21A:1-11, 2017) for the classical two-phase Lamé-Clapeyron-Stefan problem an equivalence between the temperature and convective boundary conditions at the fixed face under a certain restriction was obtained. Motivated by this article we study the two-phase Stefan problem for a semi-infinite material with a latent heat defined as a power function of the position and a convective boundary condition at the fixed face. An exact solution is constructed using Kummer functions in case that an inequality for the convective transfer coefficient is satisfied generalizing recent works for the corresponding one-phase free boundary problem. We also consider the limit to our problem when that coefficient goes to infinity obtaining a new free boundary problem, which has been recently studied in Zhou et al. (J Eng Math 2017. https://doi.org/10.1007/s10665-017-9921-y).

  11. One-dimensional two-phase thermal hydraulics (ENSTA course)

    International Nuclear Information System (INIS)

    Olive, J.

    1995-11-01

    This course is part of the ENSTA 3rd year thermal hydraulics program (nuclear power option). Its purpose is to provide the theoretical basis and main physical notions pertaining to two-phase flow, mainly focussed on water-steam flows. The introduction describes the physical specificities of these flows, emphasizing their complexity. The mathematical bases are then presented (partial derivative equations), leading to a one-dimensional type, simplified description. Balances drawn up for a pipe length volume are used to introduce the mass conservation. motion and energy equations for each phase. Various postulates used to simplify two-phase models are presented, culminating in homogeneous model definitions and equations, several common examples of which are given. The model is then applied to the calculation of pressure drops in two-phase flows. This involves presenting the models most frequently used to represent pressure drops by friction or due to pipe irregularities, without giving details (numerical values of parameters). This chapter terminates with a brief description of static and dynamic instabilities in two-phase flows. Finally, heat transfer conditions frequently encountered in liquid-steam flows are described, still in the context of a 1D model. This chapter notably includes reference to under-saturated boiling conditions and the various forms of DNB. The empirical heat transfer laws are not discussed in detail. Additional material is appended, some of which is in the form of corrected exercises. (author). 6 appends

  12. Two-phase flow characteristics analysis code: MINCS

    International Nuclear Information System (INIS)

    Watanabe, Tadashi; Hirano, Masashi; Akimoto, Masayuki; Tanabe, Fumiya; Kohsaka, Atsuo.

    1992-03-01

    Two-phase flow characteristics analysis code: MINCS (Modularized and INtegrated Code System) has been developed to provide a computational tool for analyzing two-phase flow phenomena in one-dimensional ducts. In MINCS, nine types of two-phase flow models-from a basic two-fluid nonequilibrium (2V2T) model to a simple homogeneous equilibrium (1V1T) model-can be used under the same numerical solution method. The numerical technique is based on the implicit finite difference method to enhance the numerical stability. The code structure is highly modularized, so that new constitutive relations and correlations can be easily implemented into the code and hence evaluated. A flow pattern can be fixed regardless of flow conditions, and state equations or steam tables can be selected. It is, therefore, easy to calculate physical or numerical benchmark problems. (author)

  13. BLOW-3A. A theoretical model to describe transient two-phase flow conditions in LMFBR coolant channels

    International Nuclear Information System (INIS)

    Bottoni, M.; Struwe, D.

    1982-12-01

    The computer programme BLOW-3A describes sodium boiling phenomena in subassemblies of fast breeder reactors as well as in in-pile or out-of-pile experiments simulating different failure conditions. This report presents a complete documentation of the code from three main viewpoints: the theoretical foundations of the programme are first described with particular reference to the most recent developments; the structure of the programme is then explained in all details necessary for the user to get a rapid acquaintance with it; eventually several examples of the programme validation are discussed thus enabling the reader to acquire a full picture of the possible applications of the code and at the same time to know its validity range. (orig.) [de

  14. Pressurizer and steam-generator behavior under PWR transient conditions

    International Nuclear Information System (INIS)

    Wahba, A.B.; Berta, V.T.; Pointner, W.

    1983-01-01

    Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

  15. Cold water injection into two-phase mixtures

    International Nuclear Information System (INIS)

    1989-07-01

    This report presents the results of a review of the international literature regarding the dynamic loadings associated with the injection of cold water into two-phase mixtures. The review placed emphasis on waterhammer in nuclear power plants. Waterhammmer incidence data were reviewed for information related to thermalhydraulic conditions, underlying causes and consequential damage. Condensation induced waterhammer was found to be the most significant consequence of injecting cold water into a two-phase system. Several severe waterhammer incidents have been attributed to slug formation and steam bubble collapse under conditions of stratified steam and cold water flows. These phenomena are complex and not well understood. The current body of experimental and analytical knowledge is not large enough to establish maps of expected regimes of condensation induced waterhammer. The Electric Power Research Institute, in the United States, has undertaken a major research and development programme to develop the knowledge base for this area. The limited models and data currently available show that mechanical parameters are as important as thermodynamic conditions for the initiation of condensation induced waterhammer. Examples of bounds for avoiding two-phase waterhammer are given. These bounds are system specific and depend upon parameters such as pump capacity, pipe length and pipe orientation

  16. Two phase sampling

    CERN Document Server

    Ahmad, Zahoor; Hanif, Muhammad

    2013-01-01

    The development of estimators of population parameters based on two-phase sampling schemes has seen a dramatic increase in the past decade. Various authors have developed estimators of population using either one or two auxiliary variables. The present volume is a comprehensive collection of estimators available in single and two phase sampling. The book covers estimators which utilize information on single, two and multiple auxiliary variables of both quantitative and qualitative nature. Th...

  17. CFD Simulations of Pb-Bi Two-Phase Flow

    International Nuclear Information System (INIS)

    Dostal, Vaclav; Zelezny, Vaclav; Zacha, Pavel

    2008-01-01

    In a Pb-Bi cooled direct contact steam generation fast reactor water is injected directly above the core, the produced steam is separated at the top and is send to the turbine. Neither the direct contact phenomenon nor the two-phase flow simulations in CFD have been thoroughly described yet. A first attempt in simulating such two-phase flow in 2D using the CFD code Fluent is presented in this paper. The volume of fluid explicit model was used. Other important simulation parameters were: pressure velocity relation PISO, discretization scheme body force weighted for pressure, second order upwind for momentum and CISCAM for void fraction. Boundary conditions were mass flow inlet (Pb-Bi 0 kg/s and steam 0.07 kg/s) and pressure outlet. The effect of mesh size (0.5 mm and 0.2 mm cells) was investigated as well as the effect of the turbulent model. It was found that using a fine mesh is very important in order to achieve larger bubbles and the turbulent model (k-ε realizable) is necessary to properly model the slug flow. The fine mesh and unsteady conditions resulted in computationally intense problem. This may pose difficulties in 3D simulations of the real experiments. (authors)

  18. Numerical investigation of solid-liquid two phase flow in a non-clogging centrifugal pump at off-design conditions

    International Nuclear Information System (INIS)

    Zhao, B J; Chen, H L; Hou, D H; Huang, Z F

    2012-01-01

    The solid-liquid two-phase flow fields in the non-clogging centrifugal pump with a double-channel impeller have been investigated numerically for the design condition and also off-design conditions, in order to study the solid-liquid two-phase flow pattern and non-clogging mechanism in non-clogging centrifugal pumps. The main conclusions include: The sand volume fraction distribution is extremely inhomogeneous in the whole flow channel of the pump at off-design conditions. In the impeller, particles mainly flow along the pressure surface and hub; In the volute, particles mainly accumulate in the region near to the exit of volute, the largest sand volume fraction is observed at the tongue, and a large number of particles collide with volute wall and exit the volute after circling around the volute for several times. When the particle diameter increases, particles tend to accumulate on the pressure side of the impeller, and more particles crash with the pressure side of the blade. And larger sand volume fraction gratitude is also observed in the whole flow channel of the pump. With the decrease of the inlet sand volume fraction, particles tend to accumulate on the suction side of the blade. Compared with the particle diameter, the inlet sand volume fraction has less influence on the sand volume fraction gratitude in the whole channel of the pump. At the large flow rate, the minimum and maximum sand volume fraction in the whole flow channel of the model pump tends to be smaller than that at the small flow rate. Thus, it is concluded that the water transportation capacity increases with the flow rate. This research will strengthen people's understanding of the multiphase flow pattern in non-clogging centrifugal pumps, thus provides a theoretical basis for the optimal design of non-clogging centrifugal pumps.

  19. Development and experimental evaluation of models for low capillary number two-phase flows in rough walled fractures relevant to natural gradient conditions

    International Nuclear Information System (INIS)

    Glass, R.J.; Yarrington, L.; Nicholl, M.J.

    1997-09-01

    The major results from SNL's Conceptual Model Development and Validation Task (WBS 1.2.5.4.6) as developed through exploration of small scale processes were synthesized in Glass et al. to give guidance to Performance Assessment on improving conceptual models for isothermal flow in unsaturated, fractured rock. There, pressure saturation and relative permeability curves for single fractures were proposed to be a function of both fracture orientation within the gravity field and initial conditions. We refer the reader to Glass et al. for a discussion of the implications of this behavior for Performance Assessment. The scientific research we report here substantiates this proposed behavior. We address the modeling of phase structure within fractures under natural gradient conditions relevant to unsaturated flow through fractures. This phase structure underlies the calculation of effective properties for individual fractures and hence fracture networks as required for Performance Assessment. Standard Percolation (SP) and Invasion Percolation (IP) approaches have been recently proposed to model the underlying phase saturation structures within the individual fractures during conditions of two-phase flow. Subsequent analysis of these structures yields effective two-phase pressure-saturation and relative permeability relations for the fracture. However, both of these approaches yield structures that are at odds with physical reality as we see in experiments and thus effective properties calculated from these structures are in error. Here we develop and evaluate a Modified Invasion Percolation (MIP) approach to better model quasi-static immiscible displacement in fractures. The effects of gravity, contact angle, local aperature field geometry, and local in-plane interfacial curvature between phases are included in the calculation of invasion pressure for individual sites in a discretized aperture field

  20. Experimental investigations of two-phase mixture level swell and axial void fraction distribution under high pressure, low heat flux conditions in rod bundle geometry

    International Nuclear Information System (INIS)

    Anklam, T.M.; White, M.D.

    1981-01-01

    Experimental data is reported from a series of quasi-steady-state two-phase mixture level swell and void fraction distribution tests. Testing was performed at ORNL in the Thermal Hydraulic Test Facility - a large electrically heated test loop configured to produce conditions similar to those expected in a small break loss of coolant accident. Pressure was varied from 2.7 to 8.2 MPa and linear power ranged from 0.33 to 1.95 kW/m. Mixture swell was observed to vary linearly with the total volumetric vapor generation rate over the power range of primary interest in small break analysis. Void fraction data was fit by a drift-flux model and both the drift-velocity and concentration parameter were observed to decrease with increasing pressure

  1. Uncertainty and sensitivity analysis for two-phase flow in the vicinity of the repository in the 1996 performance assessment for the Waste Isolation Pilot Plant: Disturbed conditions

    International Nuclear Information System (INIS)

    HELTON, JON CRAIG; BEAN, J.E.; ECONOMY, K.; GARNER, J.W.; MACKINNON, ROBERT J.; MILLER, JOEL D.; SCHREIBER, J.D.; VAUGHN, PALMER

    2000-01-01

    Uncertainty and sensitivity analysis results obtained in the 1996 performance assessment (PA) for the Waste Isolation Pilot Plant (WIPP) are presented for two-phase flow in the vicinity of the repository under disturbed conditions resulting from drilling intrusions. Techniques based on Latin hypercube sampling, examination of scatterplots, stepwise regression analysis, partial correlation analysis and rank transformations are used to investigate brine inflow, gas generation repository pressure, brine saturation and brine and gas outflow. Of the variables under study, repository pressure and brine flow from the repository to the Culebra Dolomite are potentially the most important in PA for the WIPP. Subsequent to a drilling intrusion repository pressure was dominated by borehole permeability and generally below the level (i.e., 8 MPa) that could potentially produce spallings and direct brine releases. Brine flow from the repository to the Culebra Dolomite tended to be small or nonexistent with its occurrence and size also dominated by borehole permeability

  2. Experiments of steady state head and torque of centrifugal pumps in two-phase flow

    International Nuclear Information System (INIS)

    Minato, Akihiko; Tominaga, Kenji.

    1988-01-01

    Circulation pump behavior has large effect on coolant discharge flow rate in case of reactor pipe break. Experiment of two-phase pump performance was conducted as a joint study of Japanese BWR user utilities and makers. Two-phase head and torque of three centrifugal pumps in high temperature and high pressure (around 6 MPa) steam/water were measured. Head was decreased from single-phase characteristics when gas was mixed in liquid flow in condition with normal flow and normal rotation directions. When flow rate was large enough, two-phase head was about the same as single-phase one in reversal flow conditions. Two-phase head was smoothly increased as flowing steam volumetic concentration increased when flow rate was small and flow direction was reversal. Changes of torque with gas concentration were correspondent to those of head. This suggested that changes of interaction between flow and impellers due to phase slip effected on torque which caused head differences between single- and two-phase flows. Dependence of dimensionless head and torque of three test pumps on steam concentration were almost the same as each other. (author)

  3. Two-phase flow in a diverging nozzle

    International Nuclear Information System (INIS)

    Wadle, M.

    1986-05-01

    Stationary two-phase flow experiments were performed with steam-water and air-water mixtures in a well-instrumented horizontal diverging nozzle. The test section consisted of a constant diameter tube, the friction-section, followed by an expansion, the diffusor, which has a tanh-contour and finally another constant diameter tube. The diameter ratio sigma=D1/D2 is 16/80. For the steam-water experiments the flow parameters were: 0 2 and for air-water mixtures (0 2 ). The initial conditions were varied to achieve subcritical and critical mass flow rates. A new model for the pressure recovery in an abrupt expansion is presented. It is based on the superficial velocity concept and agrees well with the steam-water and the water-air experimental data as well as with the experiments of other authors. The experiments were also calculated with the two-phase code DUESE. The Drift-Flux models in this code as well as the constitutive correlations and their empirical constants could be tested. It is shown, that a 1D Drift-Flux code can handle the highly transient flow in the diffusor if the proper drift model is used. In a 1D simulation it is only necessary that the computational flow area is expanded to its full width within an axial length which is equivalent to the real contour. (orig./GL) [de

  4. Wear behavior of steam generator tubes in nuclear power plant operating condition

    International Nuclear Information System (INIS)

    Kim, In-Sup; Hong, Jin-Ki; Kim, Hyung-Nam; Jang, Ki-Sang

    2003-01-01

    Reciprocating sliding wear tests were performed on steam generator tubes materials at steam generator operating temperature. The material surfaces react with oxygen to form oxides. The oxide properties such as formation rate and mechanical properties are varied with the test temperature and alloy composition. So, it is important to investigate the wear properties of each steam generator tube materials in steam generator operating condition. The tests results indicated that the wear coefficient in work rate model of alloy 690 was faster than that of alloy 800. From the scanning electron microscopy observation, the wear scars were similar each other and worn surfaces were covered with oxide layers. It seemed that the oxide layers were formed by wear debris sintering or cold welding and these layer properties affected the wear rate of steam generator tube materials. (author)

  5. The effect of steam cycle conditions upon the economics and design of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Philpott, E.F.; Pounder, F.; Willby, C.R.

    1978-01-01

    The paper studies the effect of variation of steam and feedwater conditions upon the economics, design and layout of a sodium-cooled fast reactor. The parameters investigated are steam temperature and pressure, feedwater temperature, and boiler recirculation ratio. The paper also includes an assessment of the effects of associating the fast reactor with saturated steam cycle conditions. (author)

  6. Flow instability research on steam generator with straight double-walled heat transfer tube for FBR. Pressure drop under high pressure condition

    International Nuclear Information System (INIS)

    Liu, Wei; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

    2008-01-01

    For the Steam Generator (SG) with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important items need researching. As the first step of the research, thermal hydraulics experiments were performed under high pressure condition in JAEA with using a straight tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper evaluates the pressure drop data with TRAC-BF1 code. The Pffan's correlation for single phase flow and the Martinelli-Nelson's two-phase flow multiplier are found can be well predicted the present pressure drop data under high pressure condition. (author)

  7. Condition assessment of main structural members of steam schooner WAPAMA

    Science.gov (United States)

    Xiping Wang; James Wacker; Robert Ross; Brian Brashaw

    2008-01-01

    The historic American ship WAPAMA is the last surviving example of the wooden steam-powered schooners designed for the 19th- and 20th-century Pacific Coast lumber trade and coastal service. Since its launching in 1915, the WAPAMA has had a long and productive life in plying cargo and passengers along the stormy West Coast from Mexico to Alaska. As the sole survivor of...

  8. Uniqueness of Specific Interfacial Area–Capillary Pressure–Saturation Relationship Under Non-Equilibrium Conditions in Two-Phase Porous Media Flow

    KAUST Repository

    Joekar-Niasar, Vahid

    2012-02-23

    The capillary pressure-saturation (P c-S w) relationship is one of the central constitutive relationships used in two-phase flow simulations. There are two major concerns regarding this relation. These concerns are partially studied in a hypothetical porous medium using a dynamic pore-network model called DYPOSIT, which has been employed and extended for this study: (a) P c-S w relationship is measured empirically under equilibrium conditions. It is then used in Darcy-based simulations for all dynamic conditions. This is only valid if there is a guarantee that this relationship is unique for a given flow process (drainage or imbibition) independent of dynamic conditions; (b) It is also known that P c-S w relationship is flow process dependent. Depending on drainage and imbibition, different curves can be achieved, which are referred to as "hysteresis". A thermodynamically derived theory (Hassanizadeh and Gray, Water Resour Res 29: 3389-3904, 1993a) suggests that, by introducing a new state variable, called the specific interfacial area (a nw, defined as the ratio of fluid-fluid interfacial area to the total volume of the domain), it is possible to define a unique relation between capillary pressure, saturation, and interfacial area. This study investigates these two aspects of capillary pressure-saturation relationship using a dynamic pore-network model. The simulation results imply that P c-S w relation not only depends on flow process (drainage and imbibition) but also on dynamic conditions for a given flow process. Moreover, this study attempts to obtain the first preliminary insights into the global functionality of capillary pressure-saturation-interfacial area relationship under equilibrium and non-equilibrium conditions and the uniqueness of P c-S w-a nw relationship. © 2012 The Author(s).

  9. Modelling the transition between fixed and mobile bed conditions in two-phase free-surface flows: The Composite Riemann Problem and its numerical solution

    Science.gov (United States)

    Rosatti, Giorgio; Zugliani, Daniel

    2015-03-01

    In a two-phase free-surface flow, the transition from a mobile-bed condition to a fixed-bed one (and vice versa) occurs at a sharp interface across which the relevant system of partial differential equations changes abruptly. This leads to the possibility of conceiving a new type of Riemann Problem (RP), which we have called Composite Riemann Problem (CRP), where not only the initial constant values of the variables but also the system of equations change from left to right of a discontinuity. In this paper, we present a strategy for solving a CRP by reducing it to a standard RP of a single, composite system of equations. This can be obtained by combining the two original systems by means of a suitable weighting function, namely the erodibility variable, and the introduction of an appropriate differential equation for this quantity. In this way, the CRP problem can be analyzed theoretically with standard methods, and the features of the solutions can be clearly identified. In particular, a stationary contact wave is able to correctly describe the sharp transition between mobile- and fixed-bed conditions. A finite volume scheme based on the Multiple Averages Generalized Roe approach (Rosatti and Begnudelli (2013) [22]) was used to numerically solve the fixed-mobile CRP. Several test cases demonstrate the effectiveness, exact well balanceness and high accuracy of the scheme when applied to problems that fall within the physical range of applicability of the relevant mathematical model.

  10. Two-phase flow visualization under reservoir conditions for highly heterogeneous conglomerate rock: A core-scale study for geologic carbon storage.

    Science.gov (United States)

    Kim, Kue-Young; Oh, Junho; Han, Weon Shik; Park, Kwon Gyu; Shinn, Young Jae; Park, Eungyu

    2018-03-20

    Geologic storage of carbon dioxide (CO 2 ) is considered a viable strategy for significantly reducing anthropogenic CO 2 emissions into the atmosphere; however, understanding the flow mechanisms in various geological formations is essential for safe storage using this technique. This study presents, for the first time, a two-phase (CO 2 and brine) flow visualization under reservoir conditions (10 MPa, 50 °C) for a highly heterogeneous conglomerate core obtained from a real CO 2 storage site. Rock heterogeneity and the porosity variation characteristics were evaluated using X-ray computed tomography (CT). Multiphase flow tests with an in-situ imaging technology revealed three distinct CO 2 saturation distributions (from homogeneous to non-uniform) dependent on compositional complexity. Dense discontinuity networks within clasts provided well-connected pathways for CO 2 flow, potentially helping to reduce overpressure. Two flow tests, one under capillary-dominated conditions and the other in a transition regime between the capillary and viscous limits, indicated that greater injection rates (potential causes of reservoir overpressure) could be significantly reduced without substantially altering the total stored CO 2 mass. Finally, the capillary storage capacity of the reservoir was calculated. Capacity ranged between 0.5 and 4.5%, depending on the initial CO 2 saturation.

  11. Unconditionally stable methods for simulating multi-component two-phase interface models with Peng-Robinson equation of state and various boundary conditions

    KAUST Repository

    Kou, Jisheng

    2015-03-01

    In this paper, we consider multi-component dynamic two-phase interface models, which are formulated by the Cahn-Hilliard system with Peng-Robinson equation of state and various boundary conditions. These models can be derived from the minimum problems of Helmholtz free energy or grand potential in the realistic thermodynamic systems. The resulted Cahn-Hilliard systems with various boundary conditions are fully coupled and strongly nonlinear. A linear transformation is introduced to decouple the relations between different components, and as a result, the models are simplified. From this, we further propose a semi-implicit unconditionally stable time discretization scheme, which allows us to solve the Cahn-Hilliard system by a decoupled way, and thus, our method can significantly reduce the computational cost and memory requirements. The mixed finite element methods are employed for the spatial discretization, and the approximate errors are also analyzed for both space and time. Numerical examples are tested to demonstrate the efficiency of our proposed methods. © 2015 Elsevier B.V.

  12. Analysis of reverse flow in inverted U-tubes of steam generator under natural circulation condition

    International Nuclear Information System (INIS)

    Yang Ruichang; Liu Ruolei; Liu Jinggong; Qin Shiwei

    2008-01-01

    In this paper, we report on the analysis of reverse flow in inverted U-tubes of a steam generator under natural circulation condition. The mechanism of reverse flow in inverted U-tubes of the steam generator with natural circulation is graphically analyzed by using the full-range characteristic curve of parallel U-tubes. The mathematical model and numerical calculation method for analyzing the reverse flow in inverted U-tubes of the steam generator with natural circulation have been developed. The reverse flow in an inverted U-tube steam generator of a simulated pressurized water reactor with natural circulation in analyzed. Through the calculation, the mass flow rates of normal and reverse flows in individual U-tubes are obtained. The predicted sharp drop of the fluid temperature in the inlet plenum of the steam generator due to reverse flow agrees very well with the experimental data. This indicates that the developed mathematical model and solution method can be used to correctly predict the reverse flow in the inverted U-tubes of the steam generator with natural circulation. The obtained results also show that in the analysis of natural circulation flow in the primary circuit, the reverse flow in the inverted U-tubes of the steam generator must be taken into account. (author)

  13. Uniqueness of Specific Interfacial Area–Capillary Pressure–Saturation Relationship Under Non-Equilibrium Conditions in Two-Phase Porous Media Flow

    KAUST Repository

    Joekar-Niasar, Vahid; Hassanizadeh, S. Majid

    2012-01-01

    The capillary pressure-saturation (P c-S w) relationship is one of the central constitutive relationships used in two-phase flow simulations. There are two major concerns regarding this relation. These concerns are partially studied in a

  14. Numerical simulation study of fracturing wells for shale gas with gas–water two-phase flow system under desorption and diffusion conditions

    Directory of Open Access Journals (Sweden)

    Jinzhou Zhao

    2016-06-01

    Full Text Available Hydraulic fracturing is an essential technology in developing shale gas reservoirs, not to mention, accurate prediction of productivity in fractured shale gas wells is the foundation of an efficient development in shale gas reservoirs. This paper establishes a gas–water two-phase flow percolation mathematical model by a determined numerical simulation and calculation method under desorption and diffusion conditions. By means of simulating for a post-frac performance of the shale gas reservoir, this paper devotes to a quantitative analysis the impact of fracture parameters, physical parameters, and desorption–diffusion parameters. The outcome of this research indicates that hydraulic fracturing can improve single well production and it's an effective measure in the development of shale gas. The conductivity of hydraulic fractures and the permeability of natural fractures are the main influences on shale gas production. The higher these factors are, the higher the gas and water productions are. In comparison, the matrix permeability and diffusion coefficients have minimal influences on production.

  15. CAREM-25 Steam Generator Stability Analysis

    International Nuclear Information System (INIS)

    Rabiti, A.; Delmastro, D.

    2003-01-01

    In this work the stability of a once-through CAREM-25 steam generator is analyzed.A fix nodes numerical model, that allows the modelling of the liquid, two-phase and superheated steam zones, is implemented.This model was checked against a mobile finite elements model under saturated steam conditions at the channel exit and a good agreement was obtained.Finally the stability of a CAREM steam generator is studied and the range of in let restrictions that a assure the system stability is analyzed

  16. Déplacements polyphasiques en milieu poreux. Injection de vapeur en conditions adiabatiques Multiphase Displacements in Porous Media. Steam Flooding under Adiabatic Conditions

    Directory of Open Access Journals (Sweden)

    Koci X.

    2006-11-01

    thermodynamic equilibrium is reached. The results of the simulation of the previous experiment are shown in Figs. 8 to 12. The decrease in temperature in the core (in the experiment - Fig. 6 - as well as in the simulation - Fig. 8 corresponds to a pressure decrease along the core as the steam moves towards the outlet. This decrease does not correspond to any heat losses. In fact, as the differential pressure decreases due to oil production, the temperature of the vaporization/condensation equilibrium is reduced. This induces an evolution of the residual oil saturations along the core (Fig. 12. Steam condensation at the steam front level causes an increase in water saturation just ahead (Fig. 10. The results of isothermal [1] and adiabatic displacements in terms of residual oil saturation are given in Fig. 14. The comparison is made on the basis of the temperature corresponding to the steam front. Residual oil saturations are lower under adiabatic conditions than under isothermal conditions. This is due to the fact that the mechanisms are not the same. For our experimental conditions, the characteristics of the oil (Table 3 are such that no stripping effect of the hydrocarbon phase has to be taken into account. The most important effect is assumed to be due to the steam vaporization/condensation effects. Oil recovery is a function of the spreading coefficient for the gas/oil system on the interface with the water phase [18]. One can consider that these properties are very different for a gas/water/oil system. It is then possible that phase changes can modify the displacement and hence the oil recovery. Displacement by hot water is less stable than displacement by cold water. On the contrary, displacement by steam is far less stable than displacement by hot water [19, 20] : in the case of a two-phase displacement, viscous fingering is stabilized essentially by the capillary pressure, in the case of steam injection, condensation occurs rapidly for a gas finger entering cold oil

  17. Conditions for microbial growth in the FILTRA steam absorption tower

    International Nuclear Information System (INIS)

    Nilsson, H.; Roffey, R.

    1983-08-01

    By the appointment of the Southern Sweden Power Supply an experimental study has been carried out in order to evaluate the risk for microbial growth in the planned FILTRA steam absorbtion tower at the nuclear power plant in Barsebaeck. Four modelsystems were supplied with nitrogen atmosphere and a relative humidity of 100, 75, 50 and 25 percent. The fifth system received air and 75 percent relative humidity. Samples were collected and analysed for microbial growth after 1, 2, 4 and 8 months. The amounts of microorganisms and the ATP content was monitored. No measureable growth of any significance could be observed after 8 months in any system. An elementary analyses showed that the level of nitrogen and carbon in the stones was below the limit of detection (<0.3 percent C, <0.2 percent N). (author)

  18. Two-phase flow

    International Nuclear Information System (INIS)

    Olive, J.

    1990-01-01

    The design, operation and safety of nuclear components requires increasingly accurate knowledge of two-phase flows. This knowledge is also necessary for some studies related to electricity applications. The author presents some concrete examples showing the range of problems and the complexity of the phenomena involved in these types of flows. Then, the basic principles of their numerical modelling are explained, as well as the new tendency to use increasingly local and refined models. The newest computer codes developed at EDF are briefly presented. Experimental studies dealing with twophase flow are also referred to, and their connections to numerical modelling are explained. Emphasis is placed on the major efforts devoted to the development of new test rigs and instrumentation [fr

  19. Identification of two-phase flow regimes by time-series modeling

    International Nuclear Information System (INIS)

    King, C.H.; Ouyang, M.S.; Pei, B.S.

    1987-01-01

    The identification of two-phase flow patterns in pipes or ducts is important to the design and operation of thermal-hydraulic systems, especially in the nuclear reactor cores of boiling water reactors or in the steam generators of pressurized water reactors. Basically, two-phase flow shows some fluctuating characteristics even at steady-state conditions. These fluctuating characteristics can be analyzed by statistical methods for obtaining flow signatures. There have been a number of experimental studies conducted that are concerned with the statistical properties of void fraction or pressure pulsation in two-phase flow. In this study, the authors propose a new technique of identifying the patterns of air-water two-phase flow in a vertical pipe. This technique is based on analyzing the statistic characteristics of the pressure signals of the test loop by time-series modeling

  20. Effect of Low Pressure End Conditions on Steam Power Plant Performance

    Directory of Open Access Journals (Sweden)

    Ali Syed Haider

    2014-07-01

    Full Text Available Most of the electricity produced throughout the world today is from steam power plants and improving the performance of power plants is crucial to minimize the greenhouse gas emissions and fuel consumption. Energy efficiency of a thermal power plant strongly depends on its boiler-condenser operating conditions. The low pressure end conditions of a condenser have influence on the power output, steam consumption and efficiency of a plant. Hence, the objective this paper is to study the effect of the low pressure end conditions on a steam power plant performance. For the study each component was modelled thermodynamically. Simulation was done and the results showed that performance of the condenser is highly a function of its pressure which in turn depends on the flow rate and temperature of the cooling water. Furthermore, when the condenser pressure increases both net power output and plant efficiency decrease whereas the steam consumption increases. The results can be used to run a steam power cycle at optimum conditions.

  1. Transient two-phase flow

    International Nuclear Information System (INIS)

    Hsu, Y.Y.

    1974-01-01

    The following papers related to two-phase flow are summarized: current assumptions made in two-phase flow modeling; two-phase unsteady blowdown from pipes, flow pattern in Laval nozzle and two-phase flow dynamics; dependence of radial heat and momentum diffusion; transient behavior of the liquid film around the expanding gas slug in a vertical tube; flooding phenomena in BWR fuel bundles; and transient effects in bubble two-phase flow. (U.S.)

  2. Application of Fourier transform near-infrared spectroscopy to optimization of green tea steaming process conditions.

    Science.gov (United States)

    Ono, Daiki; Bamba, Takeshi; Oku, Yuichi; Yonetani, Tsutomu; Fukusaki, Eiichiro

    2011-09-01

    In this study, we constructed prediction models by metabolic fingerprinting of fresh green tea leaves using Fourier transform near-infrared (FT-NIR) spectroscopy and partial least squares (PLS) regression analysis to objectively optimize of the steaming process conditions in green tea manufacture. The steaming process is the most important step for manufacturing high quality green tea products. However, the parameter setting of the steamer is currently determined subjectively by the manufacturer. Therefore, a simple and robust system that can be used to objectively set the steaming process parameters is necessary. We focused on FT-NIR spectroscopy because of its simple operation, quick measurement, and low running costs. After removal of noise in the spectral data by principal component analysis (PCA), PLS regression analysis was performed using spectral information as independent variables, and the steaming parameters set by experienced manufacturers as dependent variables. The prediction models were successfully constructed with satisfactory accuracy. Moreover, the results of the demonstrated experiment suggested that the green tea steaming process parameters could be predicted on a larger manufacturing scale. This technique will contribute to improvement of the quality and productivity of green tea because it can objectively optimize the complicated green tea steaming process and will be suitable for practical use in green tea manufacture. Copyright © 2011 The Society for Biotechnology, Japan. Published by Elsevier B.V. All rights reserved.

  3. Condition monitoring of steam generator by estimating the overall heat transfer coefficient

    International Nuclear Information System (INIS)

    Furusawa, Hiroaki; Gofuku, Akio

    2013-01-01

    This study develops a technique for monitoring in on-line the state of the steam generator of the fast-breeder reactor (FBR) “Monju”. Because the FBR uses liquid sodium as coolant, it is necessary to handle liquid sodium with caution due to its chemical characteristics. The steam generator generates steam by the heat of secondary sodium coolant. The sodium-water reaction may happen if a pinhole or crack occurs at the thin metal tube wall that separates the secondary sodium coolant and water/steam. Therefore, it is very important to detect an anomaly of the wall of heat transfer tubes at an early stage. This study aims at developing an on-line condition monitoring technique of the steam generator by estimating overall heat transfer coefficient from process signals. This paper describes simplified mathematical models of superheater and evaporator to estimate the overall heat transfer coefficient and a technique to diagnose the state of the steam generator. The applicability of the technique is confirmed by several estimations using simulated process signals with artificial noises. The results of the estimations show that the developed technique can detect the occurrence of an anomaly. (author)

  4. Heat transfer characteristics of horizontal steam generators under natural circulation conditions

    International Nuclear Information System (INIS)

    Hyvaerinen, J.

    1996-01-01

    This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent limitation of the decay energy removal in VVER steam generators. The limitation arises as a consequence of the steam generator tube bundle geometry. As an example, it is shown that the grace period associated with pressurizer safety valve opening during a station black-out is 2 1/2-3 hours instead of the 5-6 hours reported in several earlier studies. (However, the change in core heat-up timing is much less-about 1 h at most.) The heat transfer limitation explains the fact that, in the Greifswald VVER 440 station black-out accident in 1975, the steam generators never boiled dry. In addition, the stability of single-phase natural circulation is discussed and insights on the modelling of horizontal steam generators with general-purpose thermal-hydraulic system codes are also presented. (orig.)

  5. Flow oscillations on the steam control valve in the middle opening condition. Clarification of the phenomena by steam flow experiment and CFD calculation

    International Nuclear Information System (INIS)

    Morita, Ryo; Inada, Fumio

    2006-01-01

    A steam control valve might cause vibrations of piping when the valve opening is in a middle condition. For rationalization of maintenance and management of the plant, the valve should be improved, but it is difficult to understand flow characteristics in detail by experiment because flow around the valve is complex 3D structure and becomes supersonic (M>1). Therefore, it is necessary to clarify the cause of the vibrations and to develop improvements by Computational Fluid Dynamics (CFD) technology. In previous researches, we clarified a mechanism of the pressure fluctuations in the middle opening condition and suggested the way to prevent the pressure fluctuations by experiments and CFD calculations. But, as we used air as a working fluid in our previous research instead of steam that is used in the power plant, we couldn't consider effects of condensation and difference of change of the quantity of state between air and steam. In this report, we have conducted steam flow experiments by multi-purpose steam experiment apparatus 'WISSH' and CFD calculations by steam flow code 'MATIS-SC' to clarify those effects. As a result, in the middle opening condition, we have observed rotating pressure fluctuations in the experiment and valve-attached flow and local high-pressure region in the CFD result. These results show the pressure fluctuations in steam experiments and CFD is same kind of the fluctuations found in air experiment and CFD. (author)

  6. Uniqueness of specific interfacial area-capillary pressure-saturation relationship under non-equilibrium conditions in two-phase porous media flow

    NARCIS (Netherlands)

    Joekar-Niasar, V.; Hassanizadeh, S.M.

    2012-01-01

    The capillary pressure–saturation (P c–S w) relationship is one of the central constitutive relationships used in two-phase flow simulations. There are two major concerns regarding this relation. These concerns are partially studied in a hypothetical porous medium using a dynamic pore-network model

  7. An analysis of critical flow for steam and water extending to supercritical conditions with experimental validation

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1985-01-01

    The basic method used in this paper for establishing the critical flow of a water steam mixture including subcooled water conditions, the quality range and superheated steam conditions has already been reported and the methods are once more summarised in the next section. These methods can be extended to any fluid and results have been reported for Freon and dissociating NO/sub 2/. If an extended or complex length of pipe is involved before the position where critical flow is established, a more elaborate method is required which involves establishing the losses down the pipe. A code RAPVOID is available for analysing such cases

  8. Film boiling heat transfer from a hot sphere falling in two-phase pool

    International Nuclear Information System (INIS)

    Bang, K. H.; Kim, K. Y.

    1998-01-01

    The purpose of the present study is to experimentally investigate film boiling heat trasfer from a hot sphere falling in steam-water two-phase pool, which is the key heat transfer mode in molten fuel and coolant mixing. To measure film boiling heat transfer coefficients on a spere falling in two-phase pool, a heated sphere with a thermocouple embedded at the center is dropped in a vertical tube filled with steam-water mixture. The present experiment is unique in making the heated sphere fall through the two-phase pool while the previous experiments were performed with stationary spheres in flowing fluid. The falling speed of the sphere is measured using a set of magnet pickup coils distributed along the tube. The ranges of the experimental conditions are: spere fall speed 0-0.5 m/s, average void fraction 0-25,% steam superficial velocity 0-0.25 m/s. The results show that the forced convection film boiling heat transfer coefficient decrease slightly as the steam superficial velocity (void fraction) is increased

  9. Two-phase flow boiling pressure drop in small channels

    International Nuclear Information System (INIS)

    Sardeshpande, Madhavi V.; Shastri, Parikshit; Ranade, Vivek V.

    2016-01-01

    Highlights: • Study of typical 19 mm steam generator tube has been undertaken in detail. • Study of two phase flow boiling pressure drop, flow instability and identification of flow regimes using pressure fluctuations is the main focus of present work. • Effect of heat and mass flux on pressure drop and void fraction was studied. • Flow regimes identified from pressure fluctuations data using FFT plots. • Homogeneous model predicted pressure drop well in agreement. - Abstract: Two-phase flow boiling in small channels finds a variety of applications in power and process industries. Heat transfer, boiling flow regimes, flow instabilities, pressure drop and dry out are some of the key issues related to two-phase flow boiling in channels. In this work, the focus is on pressure drop in two-phase flow boiling in tubes of 19 mm diameter. These tubes are typically used in steam generators. Relatively limited experimental database is available on 19 mm ID tube. Therefore, in the present work, the experimental set-up is designed for studying flow boiling in 19 mm ID tube in such a way that any of the different flow regimes occurring in a steam generator tube (from pre-heating of sub-cooled water to dry-out) can be investigated by varying inlet conditions. The reported results cover a reasonable range of heat and mass flux conditions such as 9–27 kW/m 2 and 2.9–5.9 kg/m 2 s respectively. In this paper, various existing correlations are assessed against experimental data for the pressure drop in a single, vertical channel during flow boiling of water at near-atmospheric pressure. A special feature of these experiments is that time-dependent pressures are measured at four locations along the channel. The steady-state pressure drop is estimated and the identification of boiling flow regimes is done with transient characteristics using time series analysis. Experimental data and corresponding results are compared with the reported correlations. The results will be

  10. Investigations to the potential of the high temperature reactor for steam power processes with highest steam conditions and comparison with according conventional power plants

    International Nuclear Information System (INIS)

    Mondry, M.

    1988-04-01

    Already in the fifties conventional power plants with high parameters of the live steam were built to improve the total efficiency. The power plant with the highest steam conditions in the Federal Republic of Germany has 300 bar pressure and 600deg C temperature. Because of high material costs and other problems power plants with such high conditions were not continued to be built. Standard conditions of today's power plants are in the order of 180-250 bar pressure and 535deg C temperature. As the high temperature reactor is partly built up in another way than a conventional power plant, the results regarding the high steam parameters are not transferable. Possibilities for the technical realization of determined HTR-specific components are introduced and discussed. Then different HTR-power plants with steam conditions up to 350 bar pressure and 650deg C temperature are projected. Economical considerations show that an HTR with higher steam parameters brings financial profits. Further efficiency increase, which is possible by the high steam conditions, is shortly presented. The work ends with a technical and economical comparison of corresponding conventional power plants. (orig./UA) [de

  11. Comparison of steam sterilization conditions efficiency in the treatment of Infectious Health Care Waste.

    Science.gov (United States)

    Maamari, Olivia; Mouaffak, Lara; Kamel, Ramza; Brandam, Cedric; Lteif, Roger; Salameh, Dominique

    2016-03-01

    Many studies show that the treatment of Infectious Health Care Waste (IHCW) in steam sterilization devices at usual operating standards does not allow for proper treatment of Infectious Health Care Waste (IHCW). Including a grinding component before sterilization allows better waste sterilization, but any hard metal object in the waste can damage the shredder. The first objective of the study is to verify that efficient IHCW treatment can occur at standard operating parameters defined by the contact time-temperature couple in steam treatment systems without a pre-mixing/fragmenting or pre-shredding step. The second objective is to establish scientifically whether the standard operation conditions for a steam treatment system including a step of pre-mixing/fragmenting were sufficient to destroy the bacterial spores in IHCW known to be the most difficult to treat. Results show that for efficient sterilization of dialysis cartridges in a pilot 60L steam treatment system, the process would require more than 20 min at 144°C without a pre-mixing/fragmenting step. In a 720L steam treatment system including pre-mixing/fragmenting paddles, only 10 min at 144°C are required to sterilize IHCW proved to be sterilization challenges such as dialysis cartridges and diapers in normal conditions of rolling. Copyright © 2016 Elsevier Ltd. All rights reserved.

  12. Appropriate conditions for applying NaOH-pretreated two-phase olive milling waste for codigestion with food waste to enhance biogas production.

    Science.gov (United States)

    Al-Mallahi, Jumana; Furuichi, Toru; Ishii, Kazuei

    2016-02-01

    The high methane gas production potential of two phase olive milling waste (2POMW) makes its application to biogas plants in business an economical process to increase the productivity of the plants. The objective of this study was to investigate the appropriate conditions for the codigestion of NaOH-pretreated 2POMW with food waste. NaOH pretreatment can increase the methane production by increasing the soluble chemical oxygen demand (sCOD), but it may cause inhibition because of higher levels of alkalinity, sodium ion, volatile fatty acids and long chain fatty acids (LCFAs). Therefore, the first experimental phase of this study aimed to investigate the effect of different mixing ratios of 2POMW to food waste. A continuous stirred tank reactor experiment with different mixing ratios of 3%, 4.3%, 5.7% and 8.3% (2POMW: food waste) was conducted. NaOH pretreatment in the range of 6-20% was used. A mixing ratio up to 4.3%, when 10% NaOH pretreatment was used, caused no inhibition and increased methane production by 445.9mL/g-VS(2POMW). For this mixing ratio an additional experimental phase was conducted with the 20% NaOH pretreatment as the 20% NaOH pretreatment had the highest sCOD. The methane gas production was increased by 503.6mL/g-VS(2POMW). However, pH adjustment was required for applying this concentration of the high alkalinity 20% NaOH-pretreated 2POMW. Therefore, we consider using 10% NaOH pretreatment in a mixing ratio of 4.3% to be more applicable. The increase in methane gas production was correlated to the oleic acid concentration inside the reactors. The high oleic acid concentration of 61.8mg/L for the 8.3% mixing ratio was responsible for the strong inhibition. This study showed that adjusting the appropriate mixing ratio of the NaOH-pretreated 2POMW could increase the electricity production of a reactor that regularly receives food waste. Copyright © 2015 Elsevier Ltd. All rights reserved.

  13. Lumped parameter modeling of a two-phase thermal-hydraulic channel with interface tracking

    International Nuclear Information System (INIS)

    Jo, J.H.; Kaufman, J.M.; Ruger, C.J.; Stein, S.

    1978-01-01

    A nonhomogenous, thermal nonequilibrium model for one-dimensional two-phase flow in a heated channel has been formulated in lumped parameter form. The channel is divided into a variable number of flow regimes separated by moving interfaces. The model can be used to predict the behavior of a LWR core and both primary and secondary sides of a steam generator under transient conditions. (author)

  14. Optimum thermal sizing and operating conditions for once through steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Yi, Kunwoo; Ju, Kyongin; Im, Inyoung; Kim, Eunkee [KEPCO Engineering and Construction Company., Inc., Daejeon (Korea, Republic of)

    2014-10-15

    The steam generator is designed to be optimized so as to remove heat and to produce steam vapor. Because of its importance, theoretical and experimental researches have been performed on forced convection boiling heat transfer. The purpose of this study is to predict the thermal behavior and to perform optimum thermal sizing of once through steam generator. To estimate the tube thermal sizing and operating conditions of the steam generator, the analytical modeling is employed on the basis of the empirical correlation equations and theory. The optimized algorithm model, Non-dominated Sorting Genetic Algorithm (NSGA)-II, uses for this analysis. This research is focused on the design of in-vessel steam generator. An one dimensional analysis code is developed to evaluate previous researches and to optimize steam generator design parameters. The results of one-dimensional analysis need to be verified with experimental data. Goals of multi-objective optimization are to minimize tube length, pressure drop and tube number. Feedwater flow rate up to 115.425kg/s is selected so as to have margin of feedwater temperature 20 ..deg. C. For the design of 200MWth once through steam generator, it is evaluated that the tube length shall be over 12.0m for the number of tubes, 2500ea, and the length of the tube shall be over 8.0m for the number of tubes, 4500ea. The parallel coordinates chart can be provided to determine the optimal combination of number of tube, pressure drop, tube diameter and length.

  15. Study of containment air cooler capacity in steam air environment during accident conditions

    International Nuclear Information System (INIS)

    Kansal, M.; Mohan, N.; Bhawal, R.N.; Bajaj, S.S.

    2002-01-01

    Full text: The air coolers are provided for controlling the temperature in the reactor building during normal operation. These air coolers also serve as the main heat sink for the removal of energy from high enthalpy air-steam mixture expected in reactor building under accident conditions. A subroutine COOLER has been developed to estimate the heat removal rate of the air coolers at high temperature and steam conditions. The subroutine COOLER has been attached with the code PACSR (post accident containment system response) used for containment pressure temperature calculation. The subroutine was validated using design parameters at normal operating condition. A study was done to estimate the heat removal rate for some postulated accident conditions. The study reveals that, under accident conditions, the heat removal rate of air coolers increases several times compared with normal operating conditions

  16. First-principles investigations of the Ni3Sn alloy at steam reforming conditions

    DEFF Research Database (Denmark)

    Saadi, Souheil; Hinnemann, Berit; Helveg, Stig

    2009-01-01

    The structure and surface composition of a Ni3Sn alloy at conditions relevant for the steam reforming reaction was investigated using density functional theory calculations. Both the flat Ni3Sn [1 0 (1) over bar 0] surface and a surface with steps in the closed packed direction [1 0 (1) over bar 0...

  17. Entrainment in vertical annular two-phase flow

    International Nuclear Information System (INIS)

    Sawant, Pravin; Ishii, Mamoru; Mori, Michitsugu

    2009-01-01

    Prediction of amount of entrained droplets or entrainment fraction in annular two-phase flow is essential for the estimation of dryout condition and analysis of post dryout heat transfer in light water nuclear reactors and steam boilers. In this study, air-water and organic fluid (Freon-113) annular flow entrainment experiments have been carried out in 9.4 and 10.2 mm diameter test sections, respectively. Both the experiments covered three distinct pressure conditions and wide range of liquid and gas flow conditions. The organic fluid experiments simulated high pressure steam-water annular flow conditions. In each of the experiments, measurements of entrainment fraction, droplet entrainment rate and droplet deposition rate have been performed by using a liquid film extraction method. A simple, explicit and non-dimensional correlation developed by Sawant et al. (2008a) for the prediction of entrainment fraction is further improved in this study in order to account for the existence of critical gas and liquid flow rates below which no entrainment is possible. Additionally, a new correlation is proposed for the estimation of minimum liquid film flow rate at the maximum entrainment fraction condition. The improved correlation successfully predicted the newly collected air-water and Freon-113 entrainment fraction data. Furthermore, the correlations satisfactorily compared with the air-water, helium-water and air-genklene experimental data measured by Willetts (1987). (author)

  18. Flow oscillations on the steam control valve in the middle opening condition. Clarification of the effects of valve body and valve seat by steam experiments

    International Nuclear Information System (INIS)

    Morita, Ryo; Inada, Fumio

    2007-01-01

    A steam control valve might cause vibrations of piping when the valve opening is in a middle condition. For rationalization of maintenance and management of the plant, the valve should be improved, but it is difficult to understand flow characteristics in detail by experiment because flow around the valve is complex 3D structure and becomes supersonic (M>1). Therefore, it is necessary to clarify the cause of the vibrations and to develop the countermeasures by CFD (Computational Fluid Dynamics) technology. In previous researches, we clarified a mechanism of the pressure fluctuations in the middle opening condition and suggested the new valve shape (named 'Extended Valve') that can suppress the pressure fluctuations by air experiments and CFD calculations. Then, we also conducted steam experiments and CFD calculations to understand the differences between air and the steam, and found that the pressure fluctuations in the middle opening condition also occurred in the steam tests and the differences between the air and steam were not remarkable. In this report, to clarify the effects of valve and valve seat shape in steam flow condition, we conduct the steam experiments with various valve and seat shape. As a result, we find the change of the valve seat can decrease the amplitude of pressure fluctuations, but can not quite suppress the pressure fluctuations in the middle opening condition. Then, we apply the 'Extended Valve' to clarify the valve shape effect, and find that the extended valve suppresses the pressure fluctuations in the middle opening condition completely and decreases the pressure amplitude drastically. (author)

  19. Review of two-phase water hammer

    International Nuclear Information System (INIS)

    Beuthe, T.G.

    1997-01-01

    In a thermalhydraulic system like a nuclear power plant, where steam and water mix and are used to transport large amounts of energy, there is a potential to create two-phase water hammer. Large water hammer pressure transients are a threat to piping integrity and represent an important safety concern. Such events may cause unscheduled plant down time. The objective of this review is to provide a summary of the information on two-phase water hammer available in the open literature with particular emphasis on water hammer occurrences in nuclear power plants. Past reviews concentrated on studies concerned with preventing water hammer. The present review focuses on the fundamental experimental, analytical, and modelling studies. The papers discussed here were chosen from searches covering up to July 1993. (author)

  20. Modelling aspects of two phase flow

    International Nuclear Information System (INIS)

    Mayinger, F.

    1977-01-01

    In two phase flow scaling is much more limited to very narrowly defined physical phenomena than in single phase fluids. For complex and combined phenomena it can be achieved not by using dimensionless numbers alone but in addition a detailed mathematical description of the physical problem - usually in the form of a computer program - must be available. An important role plays the scaling of the thermodynamic data of the modelling fluid. From a literature survey and from own scaling experiments the conclusion can be drawn that Freon is a quite suitable modelling fluid for scaling steam-water mixtures. However, whithout a theoretical description of the phenomena nondimensional numbers for scaling two phase flow must be handled very carefully. (orig.) [de

  1. The Influence of Technological Conditions of the Process of Cogging in Flat Dies on the Quality of Two-Phase Titanium Alloys

    Directory of Open Access Journals (Sweden)

    Dyja Н.

    2016-06-01

    Full Text Available To create a rational technology of cogging process and to determinate the optimal values of the angles of tilt and single reduction the stress-strain state (SSS of the blank during cogging in the flat dies was analyzed. By using the finite element method and program MSC.SuperForge quantitative data are obtained and the basic patterns of distribution of SSS, the temperature during the simulation of tilting in flat dies with different angles of tilting and the amount of reduction were established. Sustainable experimental-industrial technology of forging of two-phase titanium alloys was developed and tested.

  2. On closure strategy for 1-D thermohydraulics models and closure relationships of two-phase flows in simple and subchannel geometry for NPP accident conditions

    International Nuclear Information System (INIS)

    Kornienko, Y.; Kornienko, E.; Ninokata, H.

    2001-01-01

    One-dimensional mathematical models are extensively used in thermohydraulics assessment of Nuclear Power Plant (NPP) transients and accidents, because specifically 1-D system of the conservation laws allows to reduce computing time and required memory, especially in ''best estimate'' code calculations. This work is generalization of the well-known Zuber-Findley and Hancox-Nicoll methods for two-phase flow distribution parameters Cs taking into account the non-monotonous void fraction distribution in the transverse direction in terms of two superimposed monotonous profiles. The method is very useful in evaluating the saddle-shape void fraction profile effects. In this work two-phase flow distribution parameters Cs were developed for simple circular and rectangular pipes, and subchannel geometry in a rod bundle. Basic assumptions were power-mode approximations for describing the profiles of local volume flux density, phase velocity and temperature. The general analytical (quadrature) relationships for Cs were obtained and their 3-D illustrations are proposed. Also, we propose generalized formulation and simple approach to construct friction factor, heat and mass transfer coefficients within the gradient hypothesis and boundary layer assumptions. The contribution of momentum, heat and mass transfer as well as their sources and sinks in the channel cross-section are taken into account. In the same way, the friction factor, heat and mass transfer coefficients with the transversal and azimuthal variations being taken into account are proposed for subchannel geometry as well. (author)

  3. Two-phase flow in refrigeration systems

    CERN Document Server

    Gu, Junjie; Gan, Zhongxue

    2013-01-01

    Two-Phase Flow in Refrigeration Systems presents recent developments from the authors' extensive research programs on two-phase flow in refrigeration systems. This book covers advanced mass and heat transfer and vapor compression refrigeration systems and shows how the performance of an automotive air-conditioning system is affected through results obtained experimentally and theoretically, specifically with consideration of two-phase flow and oil concentration. The book is ideal for university postgraduate students as a textbook, researchers and professors as an academic reference book, and b

  4. Oscillatory two-phase flows

    International Nuclear Information System (INIS)

    Boure, J.A.

    1974-12-01

    Two-phase flow instabilities are classified according to three criteria: the static or dynamic nature of the phenomenon, the necessity or not of a triggering phenomenon, and the pure or compound character of the phenomenon. Tables give the elementary instability phenomena, and the practical types of instability. Flow oscillations (or dynamic instabilities) share a number of characteristics which are dealt with, they are caused by the dynamic interactions between the flow parameters (flow rate, density, pressure, enthalpy and their distributions). Oscillation types are discussed: pure oscillations are density wave oscillations, acoustic oscillations may also occur, various compound oscillations involve either the density wave or the acoustic wave mechanism, interacting with some of the boundary conditions in the device. The analysis of slow oscillations has been made either by means of a simplified model (prediction of the thresholds) or of computer codes. Numerous computer codes are available [fr

  5. French steam generator design developments

    International Nuclear Information System (INIS)

    Ginier, R.; Campan, J.L.; Pontier, M.; Leridon, A.; Remond, A.; Castello, G.; Holcblat, A.; Paurobally, H.

    1986-01-01

    From the outset of the French nuclear power program, a significant R and D effort has been invested in improvement of the design and operation of Pressurized Water Reactors including a special committment to improving steam generators. The steam generator enhancement program has spawned a wide variety of specific R and D resources, e.g., low temperature hydraulic models for investigation of areas with single-phase flow, and freon-filled models for simulation of areas of steam generators experiencing two-phase flow (tube bundles and moisture separators). For the moisture separators, a large scale research program using freon-filled models and highly sophisticated instrumentation was used. Tests at reactor sites during startup of both 900 MWe and 1300 MWe have been used to validate the assumptions made on the basis of loop tests. These tests also demonstrated the validity of using freon to simulate two-phase flow conditions. The wealth of knowledge accumulated by the steam generator R and D program has been used to develop a new design of steam generators for the N4 plants. The current R and D effort is aimed at qualifying the N4 steam generator model and developing more comprehensive models. One prong of the R and D effort is the Megeve program. Megeve is a 25 MW steam generator which simulates operating conditions of the N4 model. The other prong is Clotaire, a freon-filled steam generator model which will be used to qualify thermal/hydraulic design codes used for multidimensional calculations for design of tube bundles

  6. Stochastic modelling of two-phase flows including phase change

    International Nuclear Information System (INIS)

    Hurisse, O.; Minier, J.P.

    2011-01-01

    Stochastic modelling has already been developed and applied for single-phase flows and incompressible two-phase flows. In this article, we propose an extension of this modelling approach to two-phase flows including phase change (e.g. for steam-water flows). Two aspects are emphasised: a stochastic model accounting for phase transition and a modelling constraint which arises from volume conservation. To illustrate the whole approach, some remarks are eventually proposed for two-fluid models. (authors)

  7. Investigation on hydrogen permeation on heat exchanger materials in conditions of steam coal gasification

    International Nuclear Information System (INIS)

    Moellenhoff, H.

    1984-01-01

    The permeation of hydrogen through iron-based alloys of different compositions in the temperature range between 700 and 1000 0 C was examined in a laboratory fluidized bed in the conditions of steam/coal gasification. Apart from tests on bright metal samples, measurement in the gasification atmosphere at a maximum pressure of 1 bar were carried out during oxidation of the metal. Experiments in a steam/hydrogen/argon mixture with the same oxidation potential were used for comparison purposes. The hydrogen permeated through the metal sample was taken to a gas chromatograph with argon flushing gas and analyzed there. The investigations on bright steel samples of various composition showed that their permeabilities for hydrogen at temperatures around 900 0 C only differed by a maximum of ± 30%. Effective prevention of permeation is therefore not possible simply by choosing a suitable alloy. If the steels are oxidized during permeation measurements, there is a reduction of the hydrogen permeability by 2 or 3 orders of magnitude due to the oxidation process, both in the steam/coal gasification fluidized bed and in a pure steam/hydrogen/argon mixture. (orig./GG) [de

  8. Condition monitoring of steam turbo generators of captive power plant at HWP (Manuguru) through vibration analysis

    International Nuclear Information System (INIS)

    Krishnareddy, G.; Chandramouli, M.; Gupta, R.V.

    2002-01-01

    Turbo Generator is a critical equipment in steam based power plant circuit. Any failure causes loss of production and hence as applicable to Heavy Water Plant, Manuguru, it results in loss of heavy water production as the captive power plant at Manuguru is solely designed to supply steam and power to Main Plant, which is meant for production of heavy water. Thereby condition monitoring is very much essential and required as part of predictive maintenance program for the turbo generators which are in continuous operation. This paper focuses on identification of the turbo generator system through vibration spectrum, characterising and differentiating the fault mechanisms, trending the faults through changes in vibration spectrums and orbit plots and subsequently planning for corrective actions/measures after evaluating the changes in machine conditions

  9. Analytical Solutions of Heat Transfer and Film Thickness with Slip Condition Effect in Thin-Film Evaporation for Two-Phase Flow in Microchannel

    Directory of Open Access Journals (Sweden)

    Ahmed Jassim Shkarah

    2015-01-01

    Full Text Available Physical and mathematical model has been developed to predict the two-phase flow and heat transfer in a microchannel with evaporative heat transfer. Sample solutions to the model were obtained for both analytical analysis and numerical analysis. It is assumed that the capillary pressure is neglected (Morris, 2003. Results are provided for liquid film thickness, total heat flux, and evaporating heat flux distribution. In addition to the sample calculations that were used to illustrate the transport characteristics, computations based on the current model were performed to generate results for comparisons with the analytical results of Wang et al. (2008 and Wayner Jr. et al. (1976. The calculated results from the current model match closely with those of analytical results of Wang et al. (2008 and Wayner Jr. et al. (1976. This work will lead to a better understanding of heat transfer and fluid flow occurring in the evaporating film region and develop an analytical equation for evaporating liquid film thickness.

  10. Study of nonequilibrium dispersed two phase flow

    International Nuclear Information System (INIS)

    Reyes, J.N. Jr.

    1986-01-01

    Understanding the behavior of liquid droplets in a superheated steam environment is essential to the accurate prediction of nuclear fuel rod surface temperatures during the blowdown and reflood phase of a loss-of-coolant-accident (LOCA). In response to this need, this treatise presents several original and significant contributions to the field of thermofluid physics. The research contained herein presents a statistical derivation of the two-phase mass, momentum, and energy-conservation equations using a droplet continuity equation analogous to that used in the Kinetic Theory of Gases. Unlike the Eulerian volume and time-averaged conservation equations generally used to describe dispersed two-phase flow behavior, this statistical averaging approach results in an additional mass momentum or energy term in each of the respective conservation equations. Further, this study demonstrates that current definitions of the volumetric vapor generation rate used in the mass conservation equation are inappropriate results under certain circumstances. The mass conservation equation derived herein is used to obtain a new definition for the volumetric vapor-generation rate. Last, a simple two phase phenomenological model, based on the statistically averaged conservation equations, is presented and solved analytically. It is shown that the actual quality and vapor temperature, under these circumstances, depend on a single dimensionless group

  11. State of the art: two-phase flow calibration techniques

    International Nuclear Information System (INIS)

    Stanley, M.L.

    1977-01-01

    The nuclear community faces a particularly difficult problem relating to the calibration of instrumentation in a two-phase flow steam/water environment. The rationale of the approach to water reactor safety questions in the United States demands that accurate measurements of mass flows in a decompressing two-phase flow be made. An accurate measurement dictates an accurate calibration. This paper addresses three questions relating to the state of the art in two-phase calibration: (1) What do we mean by calibration. (2) What is done now. (3) What should be done

  12. Predictions of structural integrity of steam generator tubes under normal operating, accident, and severe accident conditions

    International Nuclear Information System (INIS)

    Majumdar, S.

    1996-09-01

    Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating, accident, and severe accident conditions are reviewed. Tests conducted in the past, though limited, tended to show that the earlier flow-stress model for part-through-wall axial cracks overestimated the damaging influence of deep cracks. This observation is confirmed by further tests at high temperatures as well as by finite element analysis. A modified correlation for deep cracks can correct this shortcoming of the model. Recent tests have shown that lateral restraint can significantly increase the failure pressure of tubes with unsymmetrical circumferential cracks. This observation is confirmed by finite element analysis. The rate-independent flow stress models that are successful at low temperatures cannot predict the rate sensitive failure behavior of steam generator tubes at high temperatures. Therefore, a creep rupture model for predicting failure is developed and validated by tests under varying temperature and pressure loading expected during severe accidents

  13. Modelling of Zircaloy-steam-oxidation under severe fuel damage conditions

    International Nuclear Information System (INIS)

    Malang, S.; Neitzel, H.J.

    1983-01-01

    Small break loss-of-coolant accidents and special transients in an LWR, in combination with loss of required safety systems, may lead to an uncovered core for an extended period of time. As a consequence, the cladding temperature could rise up to the melting point due to the decay heat, resulting in severely damaged fuel rods. During heat-up the claddings oxidize due to oxygen uptake from the steam atmosphere in the core. The modeling and assessment of the Zircaloy-steam oxidation under such conditions is important, mainly for two reasons: The oxidation of the cladding influences the temperature transients due to the exothermic heat of reaction; the amount of liquified fuel depends on the oxide layer thickness and the oxygen content of the remaining Zircaloy metal when the melting point is reached. (author)

  14. Calculation of relative tube/tube support plate displacements in steam generators under accident condition loads using non-linear dynamic analysis methodologies

    International Nuclear Information System (INIS)

    Smith, R.E.; Waisman, R.; Hu, M.H.; Frick, T.M.

    1995-01-01

    A non-linear analysis has been performed to determine relative motions between tubes and tube support plates (TSP) during a steam line break (SLB) event for steam generators. The SLB event results in blowdown of steam and water out of the steam generator. The fluid blowdown generates pressure drops across the TSPS, resulting in out-of-plane motion. The SLB induced pressure loads are calculated with a computer program that uses a drift-flux modeling of the two-phase flow. In order to determine the relative tube/TSP motions, a nonlinear dynamic time-history analysis is performed using a structural model that considers all of the significant component members relative to the tube support system. The dynamic response of the structure to the pressure loads is calculated using a special purpose computer program. This program links the various substructures at common degrees of freedom into a combined mass and stiffness matrix. The program accounts for structural non-linearities, including potential tube and TSP interaction at any given tube position. The program also accounts for structural damping as part of the dynamic response. Incorporating all of the above effects, the equations of motion are solved to give TSP displacements at the reduced set of DOF. Using the displacement results from the dynamic analysis, plate stresses are then calculated using the detailed component models. Displacements form the dynamic analysis are imposed as boundary conditions at the DOF locations, and the finite element program then solves for the overall distorted geometry. Calculations are also performed to assure that assumptions regarding elastic response of the various structural members and support points are valid

  15. Conditions to generate Steam Fog Occurred around the Chungju Lake in the South Korea

    Science.gov (United States)

    Byungwoo, J.

    2017-12-01

    We have collected the field observation data of the steam fog occurred around the Chungju Lake in the South Korea for 3 years(2014 2016) and analyzed conditions in which the steam fog occurred. The Chungju Lake is an artificial lake made by the Chungju Dam with a water storage of 2.7 billion tons, which is the second largest in South Korea. The Chungju Dam have discharged water of the average 2.2 million tons downstream to produce electricity per day. The drainage water heats downstream of the Chungju dam and the air above water surface of downstream of that. When the warm, humid air above the downstream water mixed with cold air mass, it caused "steam fog" around the downstream of Chungju lake regardless of amount of the discharged water. The condition that promote the generation of steam fog in autumn and winter is as follows: (1) cloudless night with light winds below 1.5 m/s. (2) The differences between the temperature of discharged water from the Chungju Dam and the air temperature above the discharged water varied from 3° to 15° in autumn, from 15° to 20° in winter respectively. (3) When stream fog was generated, sensible heat flux ranged in autumn from 5 to 15 W/m2, in winter from 15 to 20 W/m2 respectively. Latent heat flux ranged in autumn from 15 to 20 W/m2, in winter from 10 to 15 W/m2 respectively.

  16. CFD Analysis of Random Turbulent Flow Load in Steam Generator of APR1400 Under Normal Operation Condition

    International Nuclear Information System (INIS)

    Lim, Sang Gyu; You, Sung Chang; Kim, Han Gon

    2011-01-01

    Regulatory guide 1.20 revision 3 of the Nuclear Regulatory Committee (NRC) modifies guidance for vibration assessments of reactor internals and steam generator internals. The new guidance requires applicants to provide a preliminary analysis and evaluation of the design and performance of a facility, including the safety margins of during normal operation and transient conditions anticipated during the life of the facility. Especially, revision 3 require rigorous assessments of adverse flow effects in the steam dryer cased by flow-excited acoustic and structural resonances such as the abnormality from power-uprated BWR cases. For two nearly identical nuclear power plants, the steam system of one BWR plant experienced failure of steam dryers and the main steam system components when steam flow was increased by 16 percent for extended power uprate (EPU). The mechanisms of those failures have revealed that a small adverse flow changing from the prototype condition induced severe flow-excited acoustic and structural resonances, leading to structural failures. In accordance with the historical background, therefore, potential adverse flow effects should be evaluated rigorously for steam generator internals in both BWR and Pressurized Water Reactor (PWR). The Advanced Power Reactor 1400 (APR1400), an evolutionary light water reactor, increased the power by 7.7 percent from the design of the 'Valid Prototype', System80+. Thus, reliable evaluations of potential adverse flow effects on the steam generator of APR1400 are necessary according to the regulatory guide. This paper is part of the computational fluid dynamics (CFD) analysis results for evaluation of the adverse flow effect for the steam generator internals of APR1400, including a series of sensitivity analyses to enhance the reliability of CFD analysis and an estimation the effect of flow loads on the internals of the steam generator under normal operation conditions

  17. Numerical simulation of two phase flows in heat exchangers

    International Nuclear Information System (INIS)

    Grandotto Biettoli, M.

    2006-04-01

    The report presents globally the works done by the author in the thermohydraulic applied to nuclear reactors flows. It presents the studies done to the numerical simulation of the two phase flows in the steam generators and a finite element method to compute these flows. (author)

  18. Two Phase Flow Split Model for Parallel Channels | Iloeje | Nigerian ...

    African Journals Online (AJOL)

    The model and code are capable of handling single and two phase flows, steady states and transients, up to ten parallel flow paths, simple and complicated geometries, including the boilers of fossil steam generators and nuclear power plants. A test calculation has been made with a simplified three-channel system ...

  19. The effect of cooling conditions on convective heat transfer and flow in a steam-cooled ribbed duct

    International Nuclear Information System (INIS)

    Shui, Linqi; Gao, Jianmin; Shi, Xiaojun; Liu, Jiazeng; Xu, Liang

    2014-01-01

    This work presents a numerical and experimental investigation on the heat transfer and turbulent flow of cooling steam in a rectangular duct with 90 .deg. ribs and studies the effect of cooling conditions on the heat transfer augmentation of steam. In the calculation, the variation range of Reynolds is from 10,000 to 190,000, the inlet temperature varies from 300 .deg. C to 500 .deg. C and the outlet pressure is from 0.5MPa to 6MPa. The aforementioned wide ranges of flow parameters cover the actual operating condition of coolant used in the gas turbine blades. The computations are carried with four turbulence models (the standard k-ε, the renormalized group (RNG) k-ε, the Launder-Reece-Rodi (LRR) and the Speziale-Sarkar-Gatski (SSG) turbulence models). The comparison of numerical and experimental results reveals that the SSG turbulence model is suitable for steam flow in the ribbed duct. Therefore, adopting the conjugate calculation technique, further study on the steam heat transfer and flow characteristics is performed with SSG turbulence model. The results show that the variation of cooling condition strongly impacts the forced convection heat transfer of steam in the ribbed duct. The cooling supply condition of a relative low temperature and medium pressure could bring a considerable advantage on steam thermal enhancement. In addition, comparing the heat transfer level between steam flow and air flow, the performance advantage of using steam is also influenced by the cooling supply condition. Changing Reynolds number has little effect on the performance superiority of steam cooling. Increasing pressure would strengthen the advantage, but increasing temperature gives an opposite result.

  20. The effect of cooling conditions on convective heat transfer and flow in a steam-cooled ribbed duct

    Energy Technology Data Exchange (ETDEWEB)

    Shui, Linqi; Gao, Jianmin; Shi, Xiaojun; Liu, Jiazeng; Xu, Liang [Xi' an Jiaotong University, Xi' an (China)

    2014-01-15

    This work presents a numerical and experimental investigation on the heat transfer and turbulent flow of cooling steam in a rectangular duct with 90 .deg. ribs and studies the effect of cooling conditions on the heat transfer augmentation of steam. In the calculation, the variation range of Reynolds is from 10,000 to 190,000, the inlet temperature varies from 300 .deg. C to 500 .deg. C and the outlet pressure is from 0.5MPa to 6MPa. The aforementioned wide ranges of flow parameters cover the actual operating condition of coolant used in the gas turbine blades. The computations are carried with four turbulence models (the standard k-ε, the renormalized group (RNG) k-ε, the Launder-Reece-Rodi (LRR) and the Speziale-Sarkar-Gatski (SSG) turbulence models). The comparison of numerical and experimental results reveals that the SSG turbulence model is suitable for steam flow in the ribbed duct. Therefore, adopting the conjugate calculation technique, further study on the steam heat transfer and flow characteristics is performed with SSG turbulence model. The results show that the variation of cooling condition strongly impacts the forced convection heat transfer of steam in the ribbed duct. The cooling supply condition of a relative low temperature and medium pressure could bring a considerable advantage on steam thermal enhancement. In addition, comparing the heat transfer level between steam flow and air flow, the performance advantage of using steam is also influenced by the cooling supply condition. Changing Reynolds number has little effect on the performance superiority of steam cooling. Increasing pressure would strengthen the advantage, but increasing temperature gives an opposite result.

  1. Study of two-phase underexpanded jets by gas jet

    International Nuclear Information System (INIS)

    Uchida, Mitsunori; Someya, Satoshi; Okamoto, Koji

    2008-01-01

    When a heat exchange in a Fast Breeder Reactor cracks, a sodium-water reaction occurs. When a tube cracks, highly pressurized water or steam escapes into the surrounding liquid sodium and a sodium-water reaction occurs forming the disodium oxide. The disodium oxide caught in the steam jet strikes other tubes in the reactor. The struck disodium oxide can then cause these tubes to crack. The release of steam into the liquid sodium media is a two-phase flow involving underexpansion. In this paper qualitative measurement of the underexpanded gas jet which injected into water was carried our for the purpose of analyzing the behavior of the two-phase flow. (author)

  2. Evaluation of steam sterilization conditions for [{sup 18}F]fludeoxyglucose

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Priscilla F.; Nascimento, Leonardo T.; Valente, Eduardo S.; Silva, Juliana B.; Silveira, Marina B.; Ferreira, Soraya Z., E-mail: somafe@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Unidade de Pesquisa e Producao de Radiofarmacos

    2011-07-01

    [{sup 18}F]Flu deoxyglucose ({sup 18}FDG) is the most commonly used radiopharmaceutical for positron emission tomography. Sterile filtration of the final product into sterile vials using 0.22 {mu}m filter membrane is usually adopted for {sup 18}FDG. However, Good Manufacturing Practice (GMP) guidelines recommend heat sterilization as the method of choice to ensure sterility of pharmaceutical preparations. The aim of this study was to essay different steam sterilization conditions in order to choose the best one for {sup 18}FDG. Three different sterilization conditions were essayed. The first one at 121 deg C for 15 minutes, the second one at 135 deg C for 3.5 minutes and the third one at 133 deg C for 2 minutes. {sup 18}FDG pH-formulation was kept around 6.0. At the end of autoclave cycles, {sup 18}FDG sterility was evaluated by direct inoculation of {sup 18}FDG in culture media and radiochemical purity was determined by TLC and HPLC. Results demonstrated that all essayed conditions were able to ensure {sup 18}FDG sterility, but caused a decrease in radiochemical purity of {sup 18}FDG. Autoclave cycle at 133 deg C for 2 minutes was the best essayed condition for {sup 18}FDG terminal sterilization, once it provided the greater radiochemical purity value and took less time. {sup 18}FDG was able to meet specifications after autoclave cycles, what supports the application of steam sterilization in routine {sup 18}FDG production, in compliance with GMP. (author)

  3. Modeling of two-phase slug flow

    International Nuclear Information System (INIS)

    Fabre, J.; Line, A.

    1992-01-01

    When gas and liquid flow in a pipe, over a range of flow rates, a flow pattern results in which sequences of long bubbles, almost filling the pipe cross section, are successively followed by liquid slugs that may contain small bubbles. This flow pattern, usually called slug flow, is encountered in numerous practical situations, such as in the production of hydrocarbons in wells and their transportation in pipelines; the production of steam and water in geothermal power plants; the boiling and condensation in liquid-vapor systems of thermal power plants; emergency core cooling of nuclear reactors; heat and mass transfer between gas and liquid in chemical reactors. This paper provides a review of two phase slug flow modeling

  4. Fictitious domain methods for elliptic problems with general boundary conditions with an application to the numerical simulation of two phase flows

    International Nuclear Information System (INIS)

    Ramiere, I.

    2006-09-01

    This work is dedicated to the introduction of two original fictitious domain methods for the resolution of elliptic problems (mainly convection-diffusion problems) with general and eventually mixed boundary conditions: Dirichlet, Robin or Neumann. The originality lies in the approximation of the immersed boundary by an approximate interface derived from the fictitious domain Cartesian mesh, which is generally not boundary-fitted to the physical domain. The same generic numerical scheme is used to impose the embedded boundary conditions. Hence, these methods require neither a surface mesh of the immersed boundary nor the local modification of the numerical scheme. We study two modelling of the immersed boundary. In the first one, called spread interface, the approximate immersed boundary is the union of the cells crossed by the physical immersed boundary. In the second one, called thin interface, the approximate immersed boundary lies on sides of mesh cells. Additional algebraic transmission conditions linking both flux and solution jumps through the thin approximate interface are introduced. The fictitious problem to solve as well as the treatment of the embedded boundary conditions are detailed for the two methods. A Q1 finite element scheme is implemented for the numerical validation of the spread interface approach while a new cell-centered finite volume scheme is derived for the thin interface approach with immersed jumps. Each method is then combined to multilevel local mesh refinement algorithms (with solution or flux residual) to increase the precision of the solution in the vicinity of the immersed interface. A convergence analysis of a Q1 finite element method with non-boundary fitted meshes is also presented. This study proves the convergence rates of the present methods. Among the various industrial applications, the simulation on a model of heat exchanger in french nuclear power plants enables us to appreciate the performances of the fictitious domain

  5. Performance evaluation of a drag-disc turbine transducer and three-beam gamma densitometer under transient two-phase flow conditions

    International Nuclear Information System (INIS)

    Nalezny, C.L.; Chen, L.L.; Solbrig, C.W.

    1979-01-01

    One of the primary variables measured in the Loss-of-Fluid Test (LOFT) Program is mass flow rate. LOFT uses drag-disc turbine tranducers (DTT) and a three-beam gamma densitometer to measure parameters from which mass flow may be computed. The transducer combination was performance tested under transient conditions in the blowdown loop at the LOFT Test Support Facility (LTSF). The performance tests consisted of three partial blowdowns of different durations starting from the same initial conditions. The reference mean mass flow rate was determined by measuring the amount of water required to reestablish initial conditions after each partial blowdown. The average mass flow rates computed from the output of the drag disc, turbine, and gamma densitometer were compared to the reference mean mass flow rates over three blowdown intervals. The tests indicated that the equal phase velocity mass measurement model provided excellent results through the use of the turbine and densitometer, and drag disc and densitometer when the phase velocities were nearly equal

  6. The significance of thermohydraulic conditions for the corrosion safety of PWR steam generators

    International Nuclear Information System (INIS)

    Gulich, J.F.

    1975-04-01

    In several PWR nuclear power plants leakages have occurred in the steam generator which were caused by localised corrosion attack. While the attention of manufacturers and operators is focused on the influences of feedwater chemistry and tube material, the present work highlights the fact that the damage always occurred in those places where flow regimed are poorly defined. The investigation leads to the result that local dry out of the heating surface can be contributing cause of damage. A method is indicated for estimating the thermohydraulic conditions in the inflow region over the tube plate and measures to improve corrosion safety are discussed. (author)

  7. Models for assessing the relative phase velocity in a two-phase flow. Status report

    International Nuclear Information System (INIS)

    Schaffrath, A.; Ringel, H.

    2000-06-01

    The knowledge of slip or drift flux in two phase flow is necessary for several technical processes (e.g. two phase pressure losses, heat and mass transfer in steam generators and condensers, dwell period in chemical reactors, moderation effectiveness of two phase coolant in BWR). In the following the most important models for two phase flow with different phase velocities (e.g. slip or drift models, analogy between pressure loss and steam quality, ε - ε models and models for the calculation of void distribution in reposing fluids) are classified, described and worked up for a further comparison with own experimental data. (orig.)

  8. Measurement of Two Phase Flow

    Directory of Open Access Journals (Sweden)

    J. Novotný

    2005-01-01

    Full Text Available This paper presents the results of experiments with moist wet steam. The aim of the experiment was to measure the velocity of the growth of a condensing nucleus in wet steam dependent on the velocity of condensation. For the experiments in wet steam an experimental setup was designed and constructed, which generated superheated steam at lowered pressure and a temperature of 50 °C. Low pressure and temperature of the hot vapour was chosen in order to minimize the risk of accidental disruption of the wall. The size of the condensing nucleus was measured by the method of Interferometric Particle Imaging (IPI. The IPI method is a technique for determining the particle size of transparent and spherical particles based on calculating the fringes captured on a CCD array. The number of fringes depends on the particle size and on the optical configuration. The experimental setup used is identical with the setup for measuring flow by the stereo PIV method. The only difference is the use of a special camera mount comprising a transparent mirror and enabling both cameras to be focused to one point. We present the results of the development of the growth of a condensing nucleus and histograms of the sizes of all measured particles depending on position and condensation velocity. 

  9. Coupling two-phase fluid flow with two-phase darcy flow in anisotropic porous media

    KAUST Repository

    Chen, J.

    2014-06-03

    This paper reports a numerical study of coupling two-phase fluid flow in a free fluid region with two-phase Darcy flow in a homogeneous and anisotropic porous medium region. The model consists of coupled Cahn-Hilliard and Navier-Stokes equations in the free fluid region and the two-phase Darcy law in the anisotropic porous medium region. A Robin-Robin domain decomposition method is used for the coupled Navier-Stokes and Darcy system with the generalized Beavers-Joseph-Saffman condition on the interface between the free flow and the porous media regions. Obtained results have shown the anisotropic properties effect on the velocity and pressure of the two-phase flow. 2014 Jie Chen et al.

  10. Coupling Two-Phase Fluid Flow with Two-Phase Darcy Flow in Anisotropic Porous Media

    Directory of Open Access Journals (Sweden)

    Jie Chen

    2014-06-01

    Full Text Available This paper reports a numerical study of coupling two-phase fluid flow in a free fluid region with two-phase Darcy flow in a homogeneous and anisotropic porous medium region. The model consists of coupled Cahn-Hilliard and Navier-Stokes equations in the free fluid region and the two-phase Darcy law in the anisotropic porous medium region. A Robin-Robin domain decomposition method is used for the coupled Navier-Stokes and Darcy system with the generalized Beavers-Joseph-Saffman condition on the interface between the free flow and the porous media regions. Obtained results have shown the anisotropic properties effect on the velocity and pressure of the two-phase flow.

  11. Development of an Enhanced Two-Phase Production System at the Geysers Geothermal Field; FINAL

    International Nuclear Information System (INIS)

    Steven Enedy

    2001-01-01

    A method was developed to enhance geothermal steam production from two-phase wells at THE Geysers Geothermal Field. The beneficial result was increased geothermal production that was easily and economically delivered to the power plant

  12. Application of a two-phase injector in the safety systems of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Popov, E; Stanev, I [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    A concept for simplification of the active part of the safety system (ASS) of nuclear power plants is presented. A two-phase injection jet device (IJD) is proposed to substitute the currently used IP-EM (impeller pumps -electric motors) couple. It is capable of sustaining a constant flow rate regardless of the variation in the system hydraulic resistance. The conditions for effective work of IJD are: development of the necessary head and flow rate, reliable supply of working medium and maintaining of the temperature of the injected water. IJD efficiency, steam and water flow rates have been calculated and compared with experimentally measured values. A short analysis of different typical accident regimes is carried out. It shows that IJD introduction brings significant advantages especially in the steam generator emergency feedwater system making it completely insensitive to loss of electricity supply accidents. 8 refs., 7 figs.

  13. Application of a two-phase injector in the safety systems of nuclear power plants

    International Nuclear Information System (INIS)

    Popov, E.; Stanev, I.

    1995-01-01

    A concept for simplification of the active part of the safety system (ASS) of nuclear power plants is presented. A two-phase injection jet device (IJD) is proposed to substitute the currently used IP-EM (impeller pumps -electric motors) couple. It is capable of sustaining a constant flow rate regardless of the variation in the system hydraulic resistance. The conditions for effective work of IJD are: development of the necessary head and flow rate, reliable supply of working medium and maintaining of the temperature of the injected water. IJD efficiency, steam and water flow rates have been calculated and compared with experimentally measured values. A short analysis of different typical accident regimes is carried out. It shows that IJD introduction brings significant advantages especially in the steam generator emergency feedwater system making it completely insensitive to loss of electricity supply accidents. 8 refs., 7 figs

  14. Temperature condition in decreasing heat transfer zone for NPP steam generators

    International Nuclear Information System (INIS)

    Kudryavtsev, I.S.; Paskar', B.L.; Sudakov, A.V.

    1985-01-01

    An experimental set-up is described and the results of temperature pulsation investigation are presented for coil steam generating channel surfaces of the NPP helium and sodium cooled HTGR. The investigations are carried out at the heat flux density of 350-900 kW/m 3 , the mass rate of 350-2000 kg/(m 2 Xs), the pressUre of 15 MPa. Temperature pulsations occur due to instability of heat transfer in the near-wall region. The results show that the critical region of burnout has a local character. Pulsation dependences on operating conditions are given. The required resource for the steam generating channel may be provided by chosing the ratio of heat flux to the mass rate, the ratio being equal to 0.5 kJ/kg for the channel with the internal diameter of 19 mm, made of the 12Kh2M steel, the wall thickness of 3 mm. In this case the maximum span of temperature pulsations doesn't exceed 25-30 K

  15. Shock wave of vapor-liquid two-phase flow

    Institute of Scientific and Technical Information of China (English)

    Liangju ZHAO; Fei WANG; Hong GAO; Jingwen TANG; Yuexiang YUAN

    2008-01-01

    The shock wave of vapor-liquid two-phase flow in a pressure-gain steam injector is studied by build-ing a mathematic model and making calculations. The results show that after the shock, the vapor is nearly com-pletely condensed. The upstream Mach number and the volume ratio of vapor have a great effect on the shock. The pressure and Mach number of two-phase shock con-form to the shock of ideal gas. The analysis of available energy shows that the shock is an irreversible process with entropy increase.

  16. Measurements of local two-phase flow parameters in a boiling flow channel

    International Nuclear Information System (INIS)

    Yun, Byong Jo; Park, Goon-CherI; Chung, Moon Ki; Song, Chul Hwa

    1998-01-01

    Local two-phase flow parameters were measured lo investigate the internal flow structures of steam-water boiling flow in an annulus channel. Two kinds of measuring methods for local two-phase flow parameters were investigated. These are a two-conductivity probe for local vapor parameters and a Pitot cube for local liquid parameters. Using these probes, the local distribution of phasic velocities, interfacial area concentration (IAC) and void fraction is measured. In this study, the maximum local void fraction in subcooled boiling condition is observed around the heating rod and the local void fraction is smoothly decreased from the surface of a heating rod to the channel center without any wall void peaking, which was observed in air-water experiments. The distributions of local IAC and bubble frequency coincide with those of local void fraction for a given area-averaged void fraction. (author)

  17. Numerical method for solution of transient, homogeneous, equilibrium, two-phase flows in one space dimension

    International Nuclear Information System (INIS)

    Shin, Y.W.; Wiedermann, A.H.

    1979-10-01

    A solution method is presented for transient, homogeneous, equilibrium, two-phase flows of a single-component fluid in one space dimension. The method combines a direct finite-difference procedure and the method of characteristics. The finite-difference procedure solves the interior points of the computing domain; the boundary information is provided by a separate procedure based on the characteristics theory. The solution procedure for boundary points requires information in addition to the physical boundary conditions. This additional information is obtained by a new procedure involving integration of characteristics in the hodograph plane. Sample problems involving various combinations of basic boundary types are calculated for two-phase water/steam mixtures and single-phase nitrogen gas, and compared with independent method-of-characteristics solutions using very fine characteristic mesh. In all cases, excellent agreement is demonstrated

  18. STUDY OF IDENTIFICATION OF TWO-PHASE FLOW PARAMETERS BY PRESSURE FLUCTUATION ANALYSIS

    Directory of Open Access Journals (Sweden)

    Ondrej Burian

    2016-12-01

    Full Text Available This paper deals with identification of parameters of simple pool boiling in a vertical rectangular channel by analysis of pressure fluctuation. In this work is introduced a small experimental facility about 9 kW power, which was used for simulation of pool boiling phenomena and creation of steam-water volume. Several pressure fluctuations measurements and differential pressure fluctuations measurements at warious were carried out. Main changed parameters were power of heaters and hydraulics resistance of channel internals. Measured pressure data was statistically analysed and compared with goal to find dependencies between parameters of two-phase flow and statistical properties of pressure fluctuation. At the end of this paper are summarized final results and applicability of this method for parameters determination of two phase flow for pool boiling conditions at ambient pressure.

  19. 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

  20. One-dimensional two-phase thermal hydraulics (ENSTA course); Thermo-hydraulique diphasique monodimensionnelle. Cours ENSTA

    Energy Technology Data Exchange (ETDEWEB)

    Olive, J

    1995-11-01

    This course is part of the ENSTA 3rd year thermal hydraulics program (nuclear power option). Its purpose is to provide the theoretical basis and main physical notions pertaining to two-phase flow, mainly focussed on water-steam flows. The introduction describes the physical specificities of these flows, emphasizing their complexity. The mathematical bases are then presented (partial derivative equations), leading to a one-dimensional type, simplified description. Balances drawn up for a pipe length volume are used to introduce the mass conservation. motion and energy equations for each phase. Various postulates used to simplify two-phase models are presented, culminating in homogeneous model definitions and equations, several common examples of which are given. The model is then applied to the calculation of pressure drops in two-phase flows. This involves presenting the models most frequently used to represent pressure drops by friction or due to pipe irregularities, without giving details (numerical values of parameters). This chapter terminates with a brief description of static and dynamic instabilities in two-phase flows. Finally, heat transfer conditions frequently encountered in liquid-steam flows are described, still in the context of a 1D model. This chapter notably includes reference to under-saturated boiling conditions and the various forms of DNB. The empirical heat transfer laws are not discussed in detail. Additional material is appended, some of which is in the form of corrected exercises. (author). 6 appends.

  1. Leak behavior of steam generator tube-to-tubesheet joints under creep condition: Experimental study

    International Nuclear Information System (INIS)

    Bahn, Chi Bum; Majumdar, Saurin; Kasza, Ken E.; Shack, William J.

    2013-01-01

    To address concerns regarding excessive leakage from throughwall cracks in steam generator tube-to-tubesheet joints under severe accident conditions, leak rate testing was conducted using tube-to-collar joint specimens. The tube interior and the interface between tube and collar (crevice) were pressurized independently using nitrogen gas. The leak rate through the crevice was almost zero when the specimens were pressurized at ∼500 °C; this low leak rate is attributed to thermal mismatch effects preventing much leakage. The near zero leak rate was maintained until the onset of large leakage at higher temperatures. The leak rate behavior after the onset of the large leakage was not much affected by the crevice length or heat-to-heat variation of Alloy 600 tubes. This suggests that once the crevice gap opens, the creep rate of the low alloy steel collar becomes dominant. Specimens with different tube diameters behaved essentially the same way. To simulate a flawed steam generator tube in the tubesheet, the crevice region was pressurized through a hole in the tube. This simulation resulted in essentially the same behavior as those specimens whose tubes and crevices were pressurized independently. Oxidation of low alloy steel collars in air tests can increase the flow resistance, and thus tests using nitrogen gas would provide more conservative leak rate data. Highlights: ► Leak rates were measured by using tube-to-collar joint specimens under creep condition. ► Leak rate through the joint interface was almost zero at ∼500 °C due to thermal mismatch. ► The near zero leak rate was maintained until the onset of large leakage at ∼680 °C. ► The leak behavior after the onset of the large leakage was not affected by hydraulic expansion length or tube heats.

  2. Research on one-dimensional two-phase flow

    International Nuclear Information System (INIS)

    Adachi, Hiromichi

    1988-10-01

    In Part I the fundamental form of the hydrodynamic basic equations for a one-dimensional two-phase flow (two-fluid model) is described. Discussions are concentrated on the treatment of phase change inertial force terms in the equations of motion and the author's equations of motion which have a remarkable uniqueness on the following three points. (1) To express force balance of unit mass two-phase fluid instead of that of unit volume two-phase fluid. (2) To pick up the unit existing mass and the unit flowing mass as the unit mass of two-phase fluid. (3) To apply the kinetic energy principle instead of the momentum low in the evaluation of steady inertial force term. In these three, the item (1) is for excluding a part of momentum change or kinetic energy change due to mass change of the examined part of fluid, which is independent of force. The item (2) is not to introduce a phenomenological physical model into the evaluation of phase change inertial force term. And the item (3) is for correctly applying the momentum law taking into account the difference of representative velocities between the main flow fluid (vapor phase or liquid phase) and the phase change part of fluid. In Part II, characteristics of various kinds of high speed two-phase flow are clarified theoretically by the basic equations derived. It is demonstrated that the steam-water two-phase critical flow with violent flashing and the airwater two-phase critical flow without phase change can be described with fundamentally the same basic equations. Furthermore, by comparing the experimental data from the two-phase critical discharge test and the theoretical prediction, the two-phase discharge coefficient, C D , for large sharp-edged orifice is determined as the value which is not affected by the experimental facility characteristics, etc. (author)

  3. Check of condition of steam generators, volume compensators and turbine condensers in nuclear power plants

    International Nuclear Information System (INIS)

    Matal, O.; Klinga, J.; Holy, F.; Sobotka, J.

    1989-01-01

    A negative pressure leak detector is described designed for leak testing of tubes in steam generators and steam turbine condensers. The principle, operation and use are described of inflatable bags and an inflatable platform. The bags are designed for insulating and sealing spaces in nuclear reactor components while the inflatable platform is used in pressurizer inspections and repairs. Their properties, and other facilities for detecting leaks in steam generator tubes are briefly described. (M.D.). 3 figs

  4. System identification on two-phase flow stability

    International Nuclear Information System (INIS)

    Wu Shaorong; Zhang Youjie; Wang Dazhong; Bo Jinghai; Wang Fei

    1996-01-01

    The theoretical principle, experimental method and results of interrelation analysis identification for the instability of two-phase flow are described. A completely new concept of test technology and method on two-phase flow stability was developed by using he theory of information science on system stability and system identification for two-phase flow stability in thermo-physics field. Application of this method would make it possible to identify instability boundary of two-phase flow under stable operation conditions of two-phase flow system. The experiment was carried out on the thermohydraulic test system HRTL-5. Using reverse repeated pseudo-random sequences of heating power as input signal sources and flow rate as response function in the test, the two-phase flow stability and stability margin of the natural circulation system are investigated. The effectiveness and feasibility of identifying two-phase flow stability by using this system identification method were experimentally demonstrated. Basic data required for mathematics modeling of two-phase flow and analysis of two-phase flow stability were obtained, which are useful for analyzing, monitoring of the system operation condition, and forecasting of two-phase flow stability in engineering system

  5. MMS two-phase nonequilibrium pressurizer

    International Nuclear Information System (INIS)

    Oh, S.J.; Sursock, J.P.

    1987-01-01

    The pressurizer of a nuclear steam supply system establishes and maintains the nuclear plant primary loop pressure within the prescribed limit. It is a vertical cylindrical vessel which provides a water reserve and a steam surge chamber to accommodate coolant density changes during operation. To adjust the pressure to a desired value, electric heaters are provided in its lower section and the spray nozzles are provided in its upper section. Also, to protect against the buildup of the excess pressure, the pressurizer has two different types of relief valves, i.e., power operated relief valve and the safety relief valve. The pressurizer model implemented to the MMS is described in detail. In particular, the handling of the nonequilibrium condition, surgeline CCFL (Counter-current Flooding Limitation), and the level tracking model are described in detail. Next, the simulation of the Shippingport pressurizer load drop test is reported

  6. Study of thermalhydraulic conditions in a steam generating channel at emergency conditions

    International Nuclear Information System (INIS)

    Staviskij, E.M.; Savvatimskij, G.I.; Elkin, I.V.; Liverant, Eh.I.; Proshutinskij, A.P.

    1980-01-01

    The results of investigation into the thermal hydraulic conditions in a ring channel when interrupting the cooling water feed are presented. Experiments have been carried out in a channel with an electrical heated inner tube of 14K18N9T steel in the range of regime parameters:Δtsub(n)=25; 100 deg C; P=3-11.9 MPa; rhow=700-2100 kg/m 2 s; q=250-850 kW/m 2 ; the temperature of the beginning of cooling tsub(p)=300-700 deg C. Analyzed are in detail Main stages of development of emergency conditions: channel dewatering and heating up, supply of a channel with water and heat transfer in the regime of film boiling, damping of heated-up wall are analyzed in detail. Formulae for the calculation of these processes and experimental data are given [ru

  7. Pre-shelling parameters and conditions that influence the whole kernel out-turn of steam-boiled cashew nuts

    Directory of Open Access Journals (Sweden)

    Babatunde Sunday Ogunsina

    2014-01-01

    Full Text Available This work investigates the effect of moisture content (MC, nut size distribution and steam exposure time (SET on the whole kernel out turn (WKO of cashew nuts during shelling using a 3 x 5 x 4 factorial experiment. Three nut sizes: small (18–22 mm, medium (23–25 mm and large (26–35 mm; five levels of MC: 8.34%, 11.80%, 12.57%, 15.40%, 16.84% (wet basis and four levels of steam exposure time (SET: 28, 30, 32, and 34 min were considered. Nuts were conditioned with warm water to the desired moisture content of 8.34%,11.80%, 12.57%, 15.40% and 16.84% (wb; and steam-boiled at 700 kPa for 28, 30,32, and 34 min. The pre-treated nuts were shelled using a hand-operated cashew nuts shelling machine. The results showed that the single effect of MC, steam exposure time (SET or nut size distribution is not enough for estimating WKO; it is rather by an interaction of these parameters. The optimum WKO of steam-boiled nuts was 91.74%, 90.94% and 87.98% for large, medium and small sized nuts at MC∗SET combination of 8.34%∗30 min, 11.80%∗32 min and 8.34%∗30 min, respectively. Pre-treatment of cashew nuts by steam boiling was found to improve whole kernel out-turn of the cashew nut. Whole kernel out-turn decreased as MC increased, thereby limiting the need for moisture adjustment when nuts are to be processed by steam boiling.

  8. A mathematical model for localized corrosion in steam generator crevices under heat transfer conditions

    International Nuclear Information System (INIS)

    Engelhardt, G.; Urquidi-Macdonald, M.; Sikora, J.; Macdonald, D.D.

    1995-01-01

    A predictive and self-consistent mathematical model has been developed to describe the localized corrosion in steam generators. The model recognizes that the internal and external environment are coupled by the need to conserve charge in the system. Thus, solution of Laplace's equation for the external environment (outside the crevice) provides the boundary condition for the electric potential at the crevice mouth, which is needed for solving the system of mass transfer equations for the internal environment (inside the crevice). Mass transfer by diffusion, ion migration, and convection was considered. Heat and momentum transfer equations are solved simultaneously, with the mass balance equation for each species and the condition of electroneutrality inside the cavity being considered. The model takes into account the porosity and tortuosity in the corrosion product deposit in the crevice. The homogeneous chemical reactions (hydrolysis of the products of the anodic reaction and the autoprotolysis of water) are included in the model. The model, in this preliminary form predicts the solution chemistry, potential drop, and temperature distribution inside the crevice. An order of magnitude estimate of the crevice corrosion rate also obtained. At this point, the model predicts only the steady state solution, but it is recognized that a steady state may not exist under normal conditions

  9. Coupling two-phase fluid flow with two-phase darcy flow in anisotropic porous media

    KAUST Repository

    Chen, J.; Sun, S.; Chen, Z.

    2014-01-01

    in the free fluid region and the two-phase Darcy law in the anisotropic porous medium region. A Robin-Robin domain decomposition method is used for the coupled Navier-Stokes and Darcy system with the generalized Beavers-Joseph-Saffman condition

  10. Experimental investigation of a two-phase nozzle flow

    International Nuclear Information System (INIS)

    Kedziur, F.; John, H.; Loeffel, R.; Reimann, J.

    1980-07-01

    Stationary two-phase flow experiments with a convergent nozzle are performed. The experimental results are appropriate to validate advanced computer codes, which are applied to the blowdown-phase of a loss-of-coolant accident (LOCA). The steam-water experiments present a broad variety of initial conditions: the pressure varies between 2 and 13 MPa, the void fraction between 0 (subcooled) and about 80%, a great number of critical as well as subcritical experiments with different flow pattern is investigated. Additional air-water experiments serve for the separation of phase transition effects. The transient acceleration of the fluid in the LOCA-case is simulated by a local acceleration in the experiment. The layout of the nozzle and the applied measurement technique allow for a separate testing of blowdown-relevant, physical models and the determination of empirical model parameters, respectively. The measured quantities are essentially the mass flow rate, quality, axial pressure and temperature profiles as well as axial and radial density/void profiles obtained by a γ-ray absorption device. Moreover, impedance probes and a pitot probe are used. Observed phenomena like a flow contraction, radial pressure and void profiles as well as the appearance of two chocking locations are described, because their examination is rather instructive about the refinement of a program. The experimental facilities as well as the data of 36 characteristic experiments are documented. (orig.) [de

  11. Temperature conditions in an LMFBR power plant from primary sodium to steam circuits

    International Nuclear Information System (INIS)

    Aubert, M.; Chaumont, J.M.; Mougniot, J.C.; Recolin, J.; Acket.

    1977-01-01

    The optimization analysis which is presented is based on an evaluation of the tender prior to contracting Super Phenix. Process constraints are reviewed: fuel limitations, turbine, steam generators; parameter selection involves major temperatures (primary ΔT 0 , steam generator water inlet temperature, turbine steam inlet temperature) or minor temperature (secondary sodium); countervailing mechanisms include upward and downward tendencies. The optimum values obtained by the method represent a coherent balanced set of parameters. So, the most significant tendency revealed by an optimization of investment costs involves the advantages of a hot system with a steam temperature above 515 0 C, but the hot temperature range is very limited (3 0 C between the hot primary sodium temperature and the steam temperature) while the cold temperatures cover a much wide range. The tolerance range within which each critical temperature may be selected without exceeding a certain cost margin per KWh is given

  12. An analytical model for prediction of two-phase (noncondensable) flow pump performance

    International Nuclear Information System (INIS)

    Furuya, O.

    1985-01-01

    During operational transients or a hypothetical LOCA (loss of coolant accident) condition, the recirculating coolant of PWR (pressurized water reactor) may flash into steam due to a loss of line pressure. Under such two-phase flow conditions, it is well known that the recirculation pump becomes unable to generate the same head as that of the single-phase flow case. Similar situations also exist in oil well submersible pumps where a fair amount of gas is contained in oil. Based on the one dimensional control volume method, an analytical method has been developed to determine the performance of pumps operating under two-phase flow conditions. The analytical method has incorporated pump geometry, void fraction, flow slippage and flow regime into the basic formula, but neglected the compressibility and condensation effects. During the course of model development, it has been found that the head degradation is mainly caused by higher acceleration on liquid phase and deceleration on gas phase than in the case of single-phase flows. The numerical results for head degradations and torques obtained with the model favorably compared with the air/water two-phase flow test data of Babcock and Wilcox (1/3 scale) and Creare (1/20 scale) pumps

  13. Two phase titanium aluminide alloy

    Science.gov (United States)

    Deevi, Seetharama C.; Liu, C. T.

    2001-01-01

    A two-phase titanic aluminide alloy having a lamellar microstructure with little intercolony structures. The alloy can include fine particles such as boride particles at colony boundaries and/or grain boundary equiaxed structures. The alloy can include alloying additions such as .ltoreq.10 at % W, Nb and/or Mo. The alloy can be free of Cr, V, Mn, Cu and/or Ni and can include, in atomic %, 45 to 55% Ti, 40 to 50% Al, 1 to 5% Nb, 0.3 to 2% W, up to 1% Mo and 0.1 to 0.3% B. In weight %, the alloy can include 57 to 60% Ti, 30 to 32% Al, 4 to 9% Nb, up to 2% Mo, 2 to 8% W and 0.02 to 0.08% B.

  14. Two-phase flow models in unbounded two-phase critical flows

    International Nuclear Information System (INIS)

    Celata, G.P.; Cumo, M.; Farello, G.E.

    1985-01-01

    With reference to a Loss-of-Coolant Accident in Light Water Reactors, an analysis of the unbounded two-phase critical flow (i.e. the issuing two-phase jet) has been accomplished. Considering jets external shape, obtained by means of photographic pictures; pressure profiles inside the jet, obtained by means of a movable ''Pitot;'' and jet phases distribution information, obtained by means of X-rays pictures; a characterization of the flow pattern in the unbounded region of a two-phase critical flow is given. Jets X-ray pictures show the existence of a central high density ''core'' gradually evaporating all around, which gives place to a characteristic ''dartflow'' the length of which depends on stagnation thermodynamic conditions

  15. Reaction kinetics of hydrazine neutralization in steam generator wet lay-up solution: Identifying optimal degradation conditions

    International Nuclear Information System (INIS)

    Schildermans, Kim; Lecocq, Raphael; Girasa, Emmanuel

    2012-09-01

    During a nuclear power plant outage, hydrazine is used as an oxygen scavenger in the steam generator lay-up solution. However, due to the carcinogenic effects of hydrazine, more stringent discharge limits are or will be imposed in the environmental permits. Hydrazine discharge could even be prohibited. Consequently, hydrazine alternatives or hydrazine degradation before discharge is needed. This paper presents the laboratory tests performed to characterize the reaction kinetics of hydrazine neutralization using bleach or hydrogen peroxide, catalyzed with either copper sulfate (CuSO 4 ) or potassium permanganate (KMnO 4 ). The tests are performed on two standard steam generator lay-up solutions based on different pH control agents: ammonia or ethanolamine. Different neutralization conditions are tested by varying temperature, oxidant addition, and catalyst concentration, among others, in order to identify the optimal parameters for hydrazine neutralization in a steam generator wet lay-up solution. (authors)

  16. Two-phase flow characterisation by nuclear magnetic resonance

    International Nuclear Information System (INIS)

    Leblond, J.; Javelot, S.; Lebrun, D.; Lebon, L.

    1998-01-01

    The results presented in this paper demonstrate the performance of the PFGSE-NMR to obtain a complete characterisation of two-phase flows. Different methods are proposed to characterise air-water flows in different regimes: stationary two-phase flows and flows in transient condition. Finally a modified PFGSE is proposed to analyse the turbulence of air-water bubbly flow. (author)

  17. Thermohydraulic verification during THTR steam generator commissioning

    International Nuclear Information System (INIS)

    Henry, C.; Elter, C.

    1988-01-01

    In one of the six THTR 300 steam generators thermocouples are installed inside the heat transfer tube bundles for measuring the gas and steam temperatures. Fluid temperature distribution measurements along and across the helix bundle have been recorded in its first months of operation over a load range of 40% up to 100% for steady state and transient conditions. Using these measurements as well as the rest of the operating instrumentation. the computer programs for the design of heat exchanger heat transfer areas are verified. The temperature measurements for steady state conditions are compared with predictions obtained in the design stage. In these codes. the heat transferred from the outside helium gas to the water/steam inside the tubes is determined in discrete steps along the heating surface by one- and two-phase heat transfer correlations. The degree of conformity between prediction and measurement is discussed and compared with more recent correlations. (author)

  18. Microgravity Two-Phase Flow Transition

    Science.gov (United States)

    Parang, M.; Chao, D.

    1999-01-01

    Two-phase flows under microgravity condition find a large number of important applications in fluid handling and storage, and spacecraft thermal management. Specifically, under microgravity condition heat transfer between heat exchanger surfaces and fluids depend critically on the distribution and interaction between different fluid phases which are often qualitatively different from the gravity-based systems. Heat transfer and flow analysis in two-phase flows under these conditions require a clear understanding of the flow pattern transition and development of appropriate dimensionless scales for its modeling and prediction. The physics of this flow is however very complex and remains poorly understood. This has led to various inadequacies in flow and heat transfer modeling and has made prediction of flow transition difficult in engineering design of efficient thermal and flow systems. In the present study the available published data for flow transition under microgravity condition are considered for mapping. The transition from slug to annular flow and from bubbly to slug flow are mapped using dimensionless variable combination developed in a previous study by the authors. The result indicate that the new maps describe the flow transitions reasonably well over the range of the data available. The transition maps are examined and the results are discussed in relation to the presumed balance of forces and flow dynamics. It is suggested that further evaluation of the proposed flow and transition mapping will require a wider range of microgravity data expected to be made available in future studies.

  19. Steam generator thermal-hydraulics

    International Nuclear Information System (INIS)

    Inch, W.W.; Scott, D.A.; Carver, M.B.

    1980-01-01

    This paper discusses a code for detailed numerical modelling of steam generator thermal-hydraulics, and describes related experimental programs designed to promote in-depth understanding of three-dimensional two-phase flow. (auth)

  20. The impact of NPP Krsko steam generator tube plugging on minimum DNBR at nominal conditions

    International Nuclear Information System (INIS)

    Lajtman, S.

    1996-01-01

    Typically, steam generator tube plugging (SGTP) both decreases the reactor coolant system (RCS) flow rate and the heat transfer surface area of the steam generator. At a constant thermal power and vessel outlet temperature, as tube plugging increases, the vessel average temperature, vessel inlet temperature and steam generator secondary side steam pressure decrease. This paper presents the analysis of impact of SGTP on Minimum Departure from Nucleate Boiling Ratio (MDNBR) at NPP Krsko (NEK), using the Improved Thermal Design Procedure (ITDP), WRB-1 correlation, and COBRA-III-C computer code. No credit was given to high plugging percentage region power reduction resulting from turbine volumetric flow limitations. MDNBR is found to be decreasing with increasing plugging, but not under the limiting values. (author)

  1. Experimental Study about Two-phase Damping Ratio on a Tube Bundle Subjected to Homogeneous Two-phase Flow

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Woo Gun; Dagdan, Banzragch [Hannam Univ., Daejeon (Korea, Republic of)

    2017-03-15

    Two-phase cross flow exists in many shell-and-tube heat exchangers such as condensers, evaporators, and nuclear steam generators. The drag force acting on a tube bundle subjected to air/water flow is evaluated experimentally. The cylinders subjected to two-phase flow are arranged in a normal square array. The ratio of pitch to diameter is 1.35, and the diameter of the cylinder is 18 mm. The drag force along the flow direction on the tube bundles is measured to calculate the drag coefficient and the two-phase damping ratio. The two-phase damping ratios, given by the analytical model for a homogeneous two-phase flow, are compared with experimental results. The correlation factor between the frictional pressure drop and the hydraulic drag coefficient is determined from the experimental results. The factor is used to calculate the drag force analytically. It is found that with an increase in the mass flux, the drag force, and the drag coefficients are close to the results given by the homogeneous model. The result shows that the damping ratio can be calculated using the homogeneous model for bubbly flow of sufficiently large mass flux.

  2. Lattice Boltzmann Methods to Address Fundamental Boiling and Two-Phase Problems

    Energy Technology Data Exchange (ETDEWEB)

    Uddin, Rizwan

    2012-01-01

    This report presents the progress made during the fourth (no cost extension) year of this three-year grant aimed at the development of a consistent Lattice Boltzmann formulation for boiling and two-phase flows. During the first year, a consistent LBM formulation for the simulation of a two-phase water-steam system was developed. Results of initial model validation in a range of thermo-dynamic conditions typical for Boiling Water Reactors (BWRs) were shown. Progress was made on several fronts during the second year. Most important of these included the simulation of the coalescence of two bubbles including the surface tension effects. Work during the third year focused on the development of a new lattice Boltzmann model, called the artificial interface lattice Boltzmann model (AILB model) for the 3 simulation of two-phase dynamics. The model is based on the principle of free energy minimization and invokes the Gibbs-Duhem equation in the formulation of non-ideal forcing function. This was reported in detail in the last progress report. Part of the efforts during the last (no-cost extension) year were focused on developing a parallel capability for the 2D as well as for the 3D codes developed in this project. This will be reported in the final report. Here we report the work carried out on testing the AILB model for conditions including the thermal effects. A simplified thermal LB model, based on the thermal energy distribution approach, was developed. The simplifications are made after neglecting the viscous heat dissipation and the work done by pressure in the original thermal energy distribution model. Details of the model are presented here, followed by a discussion of the boundary conditions, and then results for some two-phase thermal problems.

  3. Experimental CFD grade data for stratified two-phase flows

    Energy Technology Data Exchange (ETDEWEB)

    Vallee, Christophe, E-mail: c.vallee@fzd.d [Forschungszentrum Dresden-Rossendorf e.V., Institute of Safety Research, D-01314 Dresden (Germany); Lucas, Dirk; Beyer, Matthias; Pietruske, Heiko; Schuetz, Peter; Carl, Helmar [Forschungszentrum Dresden-Rossendorf e.V., Institute of Safety Research, D-01314 Dresden (Germany)

    2010-09-15

    Stratified two-phase flows were investigated at two test facilities with horizontal test-sections. For both, rectangular channel cross-sections were chosen to provide optimal observation possibilities for the application of optical measurement techniques. In order to show the local flow structure, high-speed video observation was applied, which delivers the high-resolution in space and time needed for CFD code validation. The first investigations were performed in the Horizontal Air/Water Channel (HAWAC), which is made of acrylic glass and allows the investigation of air/water co-current flows at atmospheric pressure and room temperature. At the channel inlet, a special device was designed for well-defined and adjustable inlet boundary conditions. For the quantitative analysis of the optical measurements performed at the HAWAC, an algorithm was developed to recognise the stratified interface in the camera frames. This allows to make statistical treatments for comparison with CFD calculation results. As an example, the unstable wave growth leading to slug flow is shown from the test-section inlet. Moreover, the hydraulic jump as the quasi-stationary discontinuous transition between super- and subcritical flow was investigated in this closed channel. The structure of the hydraulic jump over time is revealed by the calculation of the probability density of the water level. A series of experiments show that the hydraulic jump profile and its position from the inlet vary substantially with the inlet boundary conditions due to the momentum exchange between the phases. The second channel is built in the pressure chamber of the TOPFLOW test facility, which is used to perform air/water and steam/water experiments at pressures of up to 5.0 MPa and temperatures of up to 264 {sup o}C, but under pressure equilibrium with the vessel inside. In the present experiment, the test-section represents a flat model of the hot leg of the German Konvoi pressurised water reactor scaled at

  4. Experimental CFD grade data for stratified two-phase flows

    International Nuclear Information System (INIS)

    Vallee, Christophe; Lucas, Dirk; Beyer, Matthias; Pietruske, Heiko; Schuetz, Peter; Carl, Helmar

    2010-01-01

    Stratified two-phase flows were investigated at two test facilities with horizontal test-sections. For both, rectangular channel cross-sections were chosen to provide optimal observation possibilities for the application of optical measurement techniques. In order to show the local flow structure, high-speed video observation was applied, which delivers the high-resolution in space and time needed for CFD code validation. The first investigations were performed in the Horizontal Air/Water Channel (HAWAC), which is made of acrylic glass and allows the investigation of air/water co-current flows at atmospheric pressure and room temperature. At the channel inlet, a special device was designed for well-defined and adjustable inlet boundary conditions. For the quantitative analysis of the optical measurements performed at the HAWAC, an algorithm was developed to recognise the stratified interface in the camera frames. This allows to make statistical treatments for comparison with CFD calculation results. As an example, the unstable wave growth leading to slug flow is shown from the test-section inlet. Moreover, the hydraulic jump as the quasi-stationary discontinuous transition between super- and subcritical flow was investigated in this closed channel. The structure of the hydraulic jump over time is revealed by the calculation of the probability density of the water level. A series of experiments show that the hydraulic jump profile and its position from the inlet vary substantially with the inlet boundary conditions due to the momentum exchange between the phases. The second channel is built in the pressure chamber of the TOPFLOW test facility, which is used to perform air/water and steam/water experiments at pressures of up to 5.0 MPa and temperatures of up to 264 o C, but under pressure equilibrium with the vessel inside. In the present experiment, the test-section represents a flat model of the hot leg of the German Konvoi pressurised water reactor scaled at 1

  5. Pressure transient analysis in single and two-phase water by finite difference methods

    International Nuclear Information System (INIS)

    Berry, G.F.; Daley, J.G.

    1977-01-01

    An important consideration in the design of LMFBR steam generators is the possibility of leakage from a steam generator water tube. The ensuing sodium/water reaction will be largely controlled by the amount of water available at the leak site, thus analysis methods treating this event must have the capability of accurately modeling pressure transients through all states of water occurring in a steam generator, whether single or two-phase. The equation systems of the present model consist of the conservation equations together with an equation of state for one-dimensional homogeneous flow. These equations are then solved using finite difference techniques with phase considerations and non-equilibrium effects being treated through the equation of state. The basis for water property computation is Keenan's 'fundamental equation of state' which is applicable to single-phase water at pressures less than 1000 bars and temperatures less than 1300 0 C. This provides formulations allowing computation of any water property to any desired precision. Two-phase properties are constructed from values on the saturation line. The use of formulations permits the direct calculation of any thermodynamic property (or property derivative) to great precision while requiring very little computer storage, but does involve considerable computation time. For this reason an optional calculation scheme based on the method of 'transfinite interpolation' is included to give rapid computation in selected regions with decreased precision. The conservation equations were solved using the second order Lax-Wendroff scheme which includes wall friction, allows the formation of shocks and locally supersonic flow. Computational boundary conditions were found from a method-of-characteristics solution at the reservoir and receiver ends. The local characteristics were used to interpolate data from inside the pipe to the boundary

  6. Water spray interaction with air-steam mixtures under containment spray conditions: experimental study in the TOSQAN facility

    Energy Technology Data Exchange (ETDEWEB)

    Porcheron, E.; Lemaitre, P.; Malet, J.; Nuboer, A.; Brun, P.; Bouilloux, L.; Vendel, J. [Institut de Radioprotection et de Surete Nucleaire (IRSN), Direction de la Surete des Usines, des laboratoires, des transports et des dechets, Saclay, BP 68 - 91192 Gif-sur-Yvette cedex (France)

    2005-07-01

    Full text of publication follows: During the course of an hypothetical severe accident in a Pressurized Water Reactor (PWR), hydrogen can be produced by the reactor core oxidation and distributed into the reactor containment according to convection flows and steam wall condensation. In order to assess the risk of detonation generated by a high local hydrogen concentration, hydrogen distribution in the containment has to be known. The TOSQAN experimental program has been created to simulate typical accidental thermal hydraulic flow conditions in the reactor containment. The present work is devoted to study the interaction of a water spray injection used as a mitigation mean in order to reduce containment pressure and to produce a mixing of air, steam and hydrogen induced by spray entrainment and condensation on droplet. In order to have a better understanding of physical phenomena, we need to make a detailed characterization of the spray and the gas. The TOSQAN facility that is highly instrumented with non-intrusive diagnostics consists in a closed cylindrical vessel (7 m{sup 3} volume, 4 m high, 1.5 m i.d.) into which steam is injected. Water droplets size is measured in the vessel by the Interferometric Laser Imaging for Droplet Sizing technique. Droplet velocity is obtained by Particle Image Velocimetry and Laser Doppler Velocimetry, and droplet temperature is measured by global rainbow refractometry. Gas concentration measurements are performed by Spontaneous Raman Scattering. The walls of the vessel are thermostatically controlled by heated oil circulation. Inner spray system that is located on the top of the enclosure on the vertical axis, is composed of a single nozzle producing a full cone water spray. Spray test scenario consists of water spray injection in TOSQAN that is first pressurized with a steam injection (steam injection is stopped before spray injection). Water spray falling into the sump is removed to avoid accumulation and evaporation

  7. Water spray interaction with air-steam mixtures under containment spray conditions: comparison of heat and mass transfer modelling with the TOSQAN spray tests

    International Nuclear Information System (INIS)

    Malet, J.; Lemaitre, P.; Porcheron, E.; Vendel, J.

    2005-01-01

    using experimental results obtained in the TOSQAN facility (7 m3 volume, 4 m high, 1.5 m i.d.): presentation of this facility, its instrumentation, the spray test scenario and the experimental results are presented in a companion paper. Experimental and theoretical results are presented in two steps corresponding to the two components of this two-phase flow: one concerning the steam-air mixture, and the other concerning the water spray. Concerning the gas phase, it is shown that the vessel depressurization (and the decrease of the connected variables, such as gas temperature and humidity) during the spray injection is lower than the theoretical results. Analysis of the experimental and theoretical results is performed in order to explain this difference: the effect of the convective heat transfer between the gas and the droplets, phenomenon not taken into account in the gas modelling, is to reduce the pressure loss in the experiment. Concerning the droplet phase, droplet temperature vertical profiles downward the nozzle are presented. It is shown that the theoretical and experimental profiles are qualitatively the same, indicating a good behaviour of the heat and mass transfer modelling for the droplets. However, in order to have a quantitative agreement, droplet size has to be well determined. Sensitivity of the theoretical results to this parameter is presented and shows the importance of droplet size determination in the heat and mass transfer characterization. (authors)

  8. Steam oxidation of TP 347H FG. Laboratory exposures versus service conditions at the power plant

    Energy Technology Data Exchange (ETDEWEB)

    Hansson, Anette N. [DONG Energy A/S, Copenhagen (Denmark); Montgomery, Melanie [DONG Energy A/S, Copenhagen (Denmark); Technical Univ. of Denmark, Lyngby (Denmark). Dept. of Mechanical Engineering; Vattenfall Heat Nordic, Copenhagen (Denmark)

    2010-07-01

    TP347H FG is often used as final superheater tubing at Danish Power Plants. The oxidation behaviour of TP347H FG in steam was investigated both in laboratory conditions and field conditions. Short time exposures (336 hours) were performed in the laboratory at 500, 600 and 700 C in gasses with 8 or 46% H{sub 2}O and varying oxygen partial pressures. The shortest exposure time at the power plant was 7720 h, the temperature varied between 500 and 650 C. Surprisingly, thicker oxide layers formed within the laboratory facility at 600 and 700 C than during the long time exposures at the power plant. This could not be explained by spallation. Double-layered oxides developed during oxidation. The outer layer consist of Fe-oxides and the inner oxide contained Fe and the remaining alloy elements. Investigations with scanning electron microscopy (SEM) revealed that the morphology of the inner oxide was different for the two types of exposures. However, investigation using transmission electron microscopy (TEM) showed that the inner oxide in both cases consisted of particles of Fe-Mn-Cr spinel embedded in a metallic Fe-Ni matrix in the bulk of the (former) alloy grains and Cr-rich oxide layer along the (former) alloy grain boundaries. The main difference between the layers formed at the two locations is that the Cr-rich oxide layer is thicker for the samples exposed at the power plant than that for the samples exposed at the laboratory conditions. Furthermore, the depth of Cr depletion in the alloy adjacent the oxide layer is greater for the samples exposed at the power plant compared to those exposed in the laboratory. The microstructure investigation suggests that the slower oxidation rate of TP347H FG at the power plant as compared to the laboratory is due to a larger reservoir of Cr for the samples exposed at the power plant probably combined with a higher mobility of Cr within the alloy. (orig.)

  9. Turbulence in two-phase flows

    International Nuclear Information System (INIS)

    Sullivan, J.P.; Houze, R.N.; Buenger, D.E.; Theofanous, T.G.

    1981-01-01

    Hot film Anemometry and Laser Doppler Velocimetry have been employed in this work to study the turbulence characteristics of Bubbly and Stratified two-phase flows, respectively. Extensive consistency checks were made to establish the reliability and hence the utility of these experimental techniques for the measurement of turbulence in two-phase flows. Buoyancy-driven turbulence in vertical bubbly flows has been identified experimentally and correlated in terms of a shear velocity superposition approach. This approach provides a criterion for the demarcation of the buoyancy-driven turbulence region from the wall shear-generated turbulence region. Our data confirm the roughly isotropic behavior expected for buoyancy-driven turbulence. Upgrading of our experimental system will permit investigations of the wall-shear dominated regime (i.e., isotropy, superposition approach, etc.). The stratified flow data demonstrate clearly that the maximum in the mean velocity profile does not coincide with the zero shear plane, indicating the existence of a negative eddy viscosity region. Previous studies do not take into account this difference and thus they yield incorrect friction factor data in addition to certain puzzling behavior in the upper wall region. The conditioned turbulence data in the wavy region indicate interesting trends and that an appropriate normalization of intensities must take into account the shear velocity at the interfacial (wavy) region

  10. Macroscopic balance equations for two-phase flow models

    International Nuclear Information System (INIS)

    Hughes, E.D.

    1979-01-01

    The macroscopic, or overall, balance equations of mass, momentum, and energy are derived for a two-fluid model of two-phase flows in complex geometries. These equations provide a base for investigating methods of incorporating improved analysis methods into computer programs, such as RETRAN, which are used for transient and steady-state thermal-hydraulic analyses of nuclear steam supply systems. The equations are derived in a very general manner so that three-dimensional, compressible flows can be analysed. The equations obtained supplement the various partial differential equation two-fluid models of two-phase flow which have recently appeared in the literature. The primary objective of the investigation is the macroscopic balance equations. (Auth.)

  11. Temperature conditions in an LMFBR power plant from primary sodium to steam circuits

    International Nuclear Information System (INIS)

    Aubert, M.; Chaumont, M.; Mougniot, M.; Recolin, M.; Acket, M.

    1978-01-01

    The subject is discussed under the headings: introduction; method and procedure (optimization program, review of process constraints - fuel limitations, turbine, steam generators, parameter selection - core parameters, temperatures, counterbalancing forces affecting temperatures); optimum values of parameters, with effect on costs; conclusion, (U.K.)

  12. Lifetime forecasting of a WWER NPP steam generator tube bundle from stress corrosion conditions

    International Nuclear Information System (INIS)

    Sereda, E.V.; Gorbatykh, V.P.

    1984-01-01

    An approach is outlined to the description of corrosion cracking of austenitic stainless steels in hot chloride solutions to predict the failure of WWER NPP steam generator heat exchange tubes. The dependence of the corrosion cracking development rate on the chloride concentration and characteristic electrochemical potentials is suggsted. The approach permits also to determine the quantity of damaged tubes versus the operation parameters

  13. Thermal-hydraulics of wave propagation and pressure distribution under hypothetical steam explosion conditions in the ANS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Taleyarkhan, R.P.; Georgevich, V.; N-Valenit, S.; Kim, S.H. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    This paper describes salient aspects of the modeling and analysis framework for evaluation of dynamic loads, wave propagation, and pressure distributions (under hypothetical steam explosion conditions) around key structural boundaries of the Advanced Neutron Source (ANS) reactor core region. A staged approach was followed, using simple thermodynamic models for bounding loads and the CTH code for evaluating realistic estimates in a staged multidimensional framework. Effects of nodalization, melt dispersal into coolant during explosion, single versus multidirectional dissipation, energy level of melt, and rate of energy deposition into coolant were studied. The importance of capturing multidimensional effects that simultaneously account for fluid-structural interactions was demonstrated. As opposed to using bounding loads from thermodynamic evaluations, it was revealed that the ANS reactor system will not be vulnerable to vertically generated missiles that threaten containment if realistic estimates of energetics are used (from CTH calculations for thermally generated steam explosions without significant aluminum ignition).

  14. Two-phase-flow models and their limitations

    International Nuclear Information System (INIS)

    Ishii, M.; Kocamustafaogullari, G.

    1982-01-01

    An accurate prediction of transient two-phase flow is essential to safety analyses of nuclear reactors under accident conditions. The fluid flow and heat transfer encountered are often extremely complex due to the reactor geometry and occurrence of transient two-phase flow. Recently considerable progresses in understanding and predicting these phenomena have been made by a combination of rigorous model development, advanced computational techniques, and a number of small and large scale supporting experiments. In view of their essential importance, the foundation of various two-phase-flow models and their limitations are discussed in this paper

  15. An expert system for diagnostics and estimation of steam turbine components condition

    Science.gov (United States)

    Murmansky, B. E.; Aronson, K. E.; Brodov, Yu. M.

    2017-11-01

    The report describes an expert system of probability type for diagnostics and state estimation of steam turbine technological subsystems components. The expert system is based on Bayes’ theorem and permits to troubleshoot the equipment components, using expert experience, when there is a lack of baseline information on the indicators of turbine operation. Within a unified approach the expert system solves the problems of diagnosing the flow steam path of the turbine, bearings, thermal expansion system, regulatory system, condensing unit, the systems of regenerative feed-water and hot water heating. The knowledge base of the expert system for turbine unit rotors and bearings contains a description of 34 defects and of 104 related diagnostic features that cause a change in its vibration state. The knowledge base for the condensing unit contains 12 hypotheses and 15 evidence (indications); the procedures are also designated for 20 state parameters estimation. Similar knowledge base containing the diagnostic features and faults hypotheses are formulated for other technological subsystems of turbine unit. With the necessary initial information available a number of problems can be solved within the expert system for various technological subsystems of steam turbine unit: for steam flow path it is the correlation and regression analysis of multifactor relationship between the vibration parameters variations and the regime parameters; for system of thermal expansions it is the evaluation of force acting on the longitudinal keys depending on the temperature state of the turbine cylinder; for condensing unit it is the evaluation of separate effect of the heat exchange surface contamination and of the presence of air in condenser steam space on condenser thermal efficiency performance, as well as the evaluation of term for condenser cleaning and for tube system replacement and so forth. With a lack of initial information the expert system enables to formulate a diagnosis

  16. Transient two-phase performance of LOFT reactor coolant pumps

    International Nuclear Information System (INIS)

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed

  17. Effects of phase transformation of steam-water relative permeabilities

    Energy Technology Data Exchange (ETDEWEB)

    Verma, A.K.

    1986-03-01

    A combined theoretical and experimental study of steam-water relative permeabilities (RPs) was carried out. First, an experimental study of two-phase concurrent flow of steam and water was conducted and a set of RP curves was obtained. These curves were compared with semi-empirical and experimental results obtained by other investigators for two-phase, two-component flow (oil/gas; gas/water; gas/oil). It was found that while the wetting phase RPs were in good agreement, RPs for the steam phase were considerably higher than the non-wetting phase RPs in two-component systems. This enhancement of steam RP is attributed to phase transformation effects at the pore level in flow channels. The effects of phase transformation were studied theoretically. This study indicates that there are two separate mechanisms by which phase transformation affects RP curves: (1) Phase transformation is converging-diverging flow channels can cause an enhancement of steam phase RP. In a channel dominated by steam a fraction of the flowing steam condenses upstream from the constriction, depositing its latent heat of condensation. This heat is conducted through the solid grains around the pore throat, and evaporation takes place downstream from it. Therefore, for a given bulk flow quality; a smaller fraction of steam actually flows through the throat segments. This pore-level effect manifests itself as relative permeability enhancement on a macroscopic level; and (2) phase transformation along the interface of a stagnant phase and the phase flowing around it controls the irreducible phase saturation. Therefore, the irreducible phase saturation in steam-water flow will depend, among other factors, on the boundary conditions of the flow.

  18. Three-dimensional modeling of nuclear steam generator

    International Nuclear Information System (INIS)

    Bogdan, Z.; Afgan, N.

    1985-01-01

    In this paper mathematical model for steady-state simulation of thermodynamic and hydraulic behaviour of U-tube nuclear steam generator is described. The model predicts three-dimensional distribution of temperatures, pressures, steam qualities and velocities in the steam generator secondary loop. In this analysis homogeneous two phase flow model is utilized. Foe purpose of the computer implementation of the mathematical model, a special flow distribution code NUGEN was developed. Calculations are performed with the input data and geometrical characteristics related to the D-4 (westinghouse) model of U-tube nuclear steam generator built in Krsko, operating under 100% load conditions. Results are shown in diagrams giving spatial distribution of pertinent variables in the secondary loop. (author)

  19. Computer simulation of the steam--graphite reaction under isothermal and steady-state conditions

    International Nuclear Information System (INIS)

    Joy, D.S.; Stem, S.C.

    1975-05-01

    A mathematical model was formulated to describe the isothermal, steady-state diffusion and reaction of steam in a graphite matrix. A generalized Langmuir-Hinshelwood equation is used to represent the steam-graphite reaction rate. The model also includes diffusion in the gas phase adjacent to the graphite matrix. A computer program, written to numerically integrate the resulting differential equations, is described. The coupled nonlinear differential equations in the graphite phase are solved using the IBM Continuous System Modeling Program. Classical finite difference techniques are used for the gas-phase calculations. An iterative procedure is required to couple the two sets of calculations. Several sample problems are presented to demonstrate the utility of the model. (U.S.)

  20. IMPACT OF THE COLD END OPERATING CONDITIONS ON ENERGY EFFICIENCY OF THE STEAM POWER PLANTS

    Directory of Open Access Journals (Sweden)

    Slobodan Laković

    2010-01-01

    Full Text Available The conventional steam power plant working under the Rankine Cycle and the steam condenser as a heat sink and the steam boiler as a heat source have the same importance for the power plant operating process. Energy efficiency of the coal fired power plant strongly depends on its turbine-condenser system operation mode. For the given thermal power plant configuration, cooling water temperature or/and flow rate change generate alterations in the condenser pressure. Those changes have great influence on the energy efficiency of the plant. This paper focuses on the influence of the cooling water temperature and flow rate on the condenser performance, and thus on the specific heat rate of the coal fired plant and its energy efficiency. Reference plant is working under turbine-follow mode with an open cycle cooling system. Analysis is done using thermodynamic theory, in order to define heat load dependence on the cooling water temperature and flow rate. Having these correlations, for given cooling water temperature it is possible to determine optimal flow rate of the cooling water in order to achieve an optimal condensing pressure, and thus, optimal energy efficiency of the plant. Obtained results could be used as useful guidelines in improving existing power plants performances and also in design of the new power plants.

  1. Experimental and Analytical Study of Lead-Bismuth-Water Direct Contact Boiling Two-Phase Flow

    Science.gov (United States)

    Novitrian; Dostal, Vaclav; Takahashi, Minoru

    The characteristics of lead-bismuth(Pb-Bi)-water boiling two-phase flow were investigated experimentally and analytically using a Pb-Bi-water direct contact boiling two-phase flow loop. Pb-Bi flow rates and void fraction were measured in a vertical circular tube at conditions of system pressure 7MPa, liquid metal temperature 460°C and injected water temperature 220°C. The drift-flux model with the assumption that bubble sizes were dependent on the fluid surface tension and the density ratio of Pb-Bi to steam-water mixture was chosen and modified by the best fit to the measured void fraction. Pb-Bi flow rates were analytically estimated using balance condition between buoyancy force and pressure losses, where the buoyancy force was calculated from void fraction estimated using the modified drift-flux model. The deviation of the analytical results of the flow rates from the experimental ones was less than 10%.

  2. Interfacial Instability in Two-Phase Flow: Manipulating Coalescence and Condensation

    Data.gov (United States)

    National Aeronautics and Space Administration — Two-phase flow under microgravity conditions presents a number of technical challenges ( and ). Life support and habitation depend on systems that use two-phase flow...

  3. Two-phase flow measurements with advanced instrumented spool pieces and local conductivity probes

    International Nuclear Information System (INIS)

    Turnage, K.G.; Davis, C.E.

    1979-01-01

    A series of two-phase, air-water and steam-water tests performed with instrumented spool pieces and with conductivity probes obtained from Atomic Energy of Canada, Ltd. is described. The behavior of the three-beam densitometer, turbine meter, and drag flowmeter is discussed in terms of two-phase models. Application of some two-phase mass flow models to the recorded spool piece data is made and preliminary results are shown. Velocity and void fraction information derived from the conductivity probes is presented and compared to velocities and void fractions obtained using the spool piece instrumentation

  4. One-dimensional transient unequal velocity two-phase flow by the method of characteristics

    International Nuclear Information System (INIS)

    Rasouli, F.

    1981-01-01

    An understanding of two-phase flow is important when one is analyzing the accidental loss of coolant or when analyzing industrial processes. If a pipe in the steam generator of a nuclear reactor breaks, the flow will remain critical (or choked) for almost the entire blowdown. For this reason the knowledge of the two-phase maximum (critical) flow rate is important. A six-equation model--consisting of two continuity equations, two energy equations, a mixture momentum equation, and a constitutive relative velocity equation--is solved numerically by the method of characteristics for one-dimensional, transient, two-phase flow systems. The analysis is also extended to the special case of transient critical flow. The six-equation model is used to study the flow of a nonequilibrium sodium-argon system in a horizontal tube in which the nonequilibrium sodium-argon system in a horizontal tube in which the critical flow condition is at the entrance. A four-equation model is used to study the pressure-pulse propagation rate in an isothermal air-water system, and the results that are found are compared with the experimental data. Proper initial and boundary conditions are obtained for the blowdown problem. The energy and mass exchange relations are evaluated by comparing the model predictions with results of void-fraction and heat-transfer experiments. A simplified two-equation model is obtained for the special case of two incompressible phases. This model is used in the preliminary analysis of batch sedimentation. It is also used to predict the shock formation in the gas-solid fluidized bed

  5. Non-contact control of the working condition of mechanical units of the steam compressor for desalination plant

    Science.gov (United States)

    Danilin, A. I.; Chernyavsky, A. Zh.; Danilin, S. A.; Neverov, V. V.; Voroh, D. A.; Blagin, E. V.

    2018-03-01

    New methods and means for monitoring working condition of the rotating elements of steam compressor unit such as blade ring of the impeller and gears of multiplier are considered. Blade control is carried out by the signalling device of pre-emergency deformation of impeller blades. Control of the gears condition is carried out by apparatus system which allows to analyse change of the signal form caused by the gears wear. Influence of the wear types on the typical information parameters of the analysed signals is described. Technical characteristics of the devices and experimental research results are presented. Described control systems allow to detect deviations equal to 1-2% from initial condition. Application of such systems gives the opportunity to improve fault diagnosis and maintenance in 2-3 times.

  6. Steam sterilization does not require saturated steam

    NARCIS (Netherlands)

    van Doornmalen Gomez Hoyos, J. P.C.M.; Paunovic, A.; Kopinga, K.

    2017-01-01

    The most commonly applied method to sterilize re-usable medical devices in hospitals is steam sterilization. The essential conditions for steam sterilization are derived from sterilization in water. Microbiological experiments in aqueous solutions have been used to calculate various time–temperature

  7. The development of two-phase flow instrumentation at PNC O-arai Engineering Center

    International Nuclear Information System (INIS)

    Obata, T.; Kobori, T.; Hayamizu, Y.

    1975-10-01

    This paper reviews development works on the two-phase flow instrumentation carried out at PNC Oarai Engineering Center for FUGEN safety test. The paper describes heater surface temperature measurement, four types of void meters and two steam quality meters. (auth.)

  8. AMPTRACT: an algebraic model for computing pressure tube circumferential and steam temperature transients under stratified channel coolant conditions

    International Nuclear Information System (INIS)

    Gulshani, P.; So, C.B.

    1986-10-01

    In a number of postulated accident scenarios in a CANDU reactor, some of the horizontal fuel channels are predicted to experience periods of stratified channel coolant condition which can lead to a circumferential temperature gradient around the pressure tube. To study pressure tube strain and integrity under stratified flow channel conditions, it is, necessary to determine the pressure tube circumferential temperature distribution. This paper presents an algebraic model, called AMPTRACT (Algebraic Model for Pressure Tube TRAnsient Circumferential Temperature), developed to give the transient temperature distribution in a closed form. AMPTRACT models the following modes of heat transfer: radiation from the outermost elements to the pressure tube and from the pressure to calandria tube, convection between the fuel elements and the pressure tube and superheated steam, and circumferential conduction from the exposed to submerged part of the pressure tube. An iterative procedure is used to solve the mass and energy equations in closed form for axial steam and fuel-sheath transient temperature distributions. The one-dimensional conduction equation is then solved to obtain the pressure tube circumferential transient temperature distribution in a cosine series expansion. In the limit of large times and in the absence of convection and radiation to the calandria tube, the predicted pressure tube temperature distribution reduces identically to a parabolic profile. In this limit, however, radiation cannot be ignored because the temperatures are generally high. Convection and radiation tend to flatten the parabolic distribution

  9. Effects of raw material extrusion and steam conditioning on feed pellet quality and nutrient digestibility of growing meat rabbits.

    Science.gov (United States)

    Liao, Kuoyao; Cai, Jingyi; Shi, Zhujun; Tian, Gang; Yan, Dong; Chen, Delin

    2017-06-01

    This study was conducted to investigate the effects of raw material extrusion and steam conditioning on feed pellet quality and nutrient digestibility of growing meat rabbits, in order to determine appropriate rabbit feed processing methods and processing parameters. In Exp. 1, an orthogonal design was adopted. Barrel temperature, material moisture content and feed rate were selected as test factors, and acid detergent fiber (ADF) content was selected as an evaluation index to research the optimum extrusion parameters. In Exp. 2, a two-factor design was adopted. Four kinds of rabbit feeds were processed and raw material extrusion adopted optimum extrusion parameters of Exp. 1. A total of 40 healthy and 42-day-old rabbits with similar weight were used in a randomized design, which consisted of 4 groups and 10 replicates in each group (1 rabbits in each replicate). The adaptation period lasted for 7 d, and the digestion trial lasted for 4 d. The results showed as follows: 1) ADF was significantly affected by barrel temperature ( P  digestibility of dry matter and total energy ( P  digestibility of crude fiber (CF), ADF and NDF ( P  digestibility of rabbit feed. Thus, using extrusion and steam conditioning technology at the same time in the weaning rabbits feed processing can improve the pellet quality and nutrient apparent digestibility of rabbit feed.

  10. Two phases of droplet evaporation during plasma arc spraying: reply to Chen's comment about the 'rocket' effect under conditions of thermal plasma spraying

    Energy Technology Data Exchange (ETDEWEB)

    Nemchinsky, V A [Key College, 225 Dania Beach Blvd, Dania Beach, FL 33040 (United States)

    2007-07-07

    The heating history of a droplet during its flight can be divided into two phases: (a) the initial phase when evaporation, although it occurs, does not change the heat balance of the droplet much (the case considered in our previous paper and (b) the final phase when the cooling due to evaporation balances the heat flux from the plasma. The later phase is considered in Chen's comment. In our reply, a very straightforward consideration demonstrates that even in the final phase of the droplet flight, the 'rocket' effect can be significant. (reply)

  11. An experimental study of two-phase natural circulation in an adiabatic flow loop

    International Nuclear Information System (INIS)

    Tan, M.J.; Lambert, G.A.; Ishii, Mamoru.

    1988-01-01

    An experimental investigation was conducted to study the two-phase flow aspect of the phenomena of interruption and resumption of natural circulation, two-phase flow patterns and pattern transitions in the hot legs of B and W light water reactor systems. The test facility was a scaled adiabatic loop designed in accordance with the scaling criteria developed by Kocamustafaogullari and Ishii. The diameter and the height of the hot leg were 10 cm and 5.5 m, respectively; the working fluid pair was nitrogen-water. The effects of the thermal center in the steam generators, friction loss in the cold leg, and configuration of the inlet to the hot leg on the flow conditions in the hot leg were investigated by varying the water level in a gas separator, controlling the size of opening of a friction loss control valve, and using two inlet geometries. Methods for estimating the distribution parameter and the average drift velocity are proposed so that they may be used in the application of one-dimensional drift-flux model to the analysis of the interruption and resumption of natural circulation in a similar geometry. 7 refs., 17 figs., 4 tabs

  12. Advanced numerical methods for three dimensional two-phase flow calculations in PWR

    International Nuclear Information System (INIS)

    Toumi, I.; Gallo, D.; Royer, E.

    1997-01-01

    This paper is devoted to new numerical methods developed for three dimensional two-phase flow calculations. These methods are finite volume numerical methods. They are based on an extension of Roe's approximate Riemann solver to define convective fluxes versus mean cell quantities. To go forward in time, a linearized conservative implicit integrating step is used, together with a Newton iterative method. We also present here some improvements performed to obtain a fully implicit solution method that provides fast running steady state calculations. This kind of numerical method, which is widely used for fluid dynamic calculations, is proved to be very efficient for the numerical solution to two-phase flow problems. This numerical method has been implemented for the three dimensional thermal-hydraulic code FLICA-4 which is mainly dedicated to core thermal-hydraulic transient and steady-state analysis. Hereafter, we will also find some results obtained for the EPR reactor running in a steady-state at 60% of nominal power with 3 pumps out of 4, and a thermal-hydraulic core analysis for a 1300 MW PWR at low flow steam-line-break conditions. (author)

  13. Two-phased flow component loss data

    International Nuclear Information System (INIS)

    Fairhurst, C.P.

    1983-01-01

    Pressure loss measurements were made for valves and orifice plates under horizontal and vertical two-phase, air/water flow. The results displayed similar trends and were successfully correlated using a semi-empirical approach. (author)

  14. Constitutive equations for two-phase flows

    International Nuclear Information System (INIS)

    Boure, J.A.

    1974-12-01

    The mathematical model of a system of fluids consists of several kinds of equations complemented by boundary and initial conditions. The first kind equations result from the application to the system, of the fundamental conservation laws (mass, momentum, energy). The second kind equations characterize the fluid itself, i.e. its intrinsic properties and in particular its mechanical and thermodynamical behavior. They are the mathematical model of the particular fluid under consideration, the laws they expressed are so called the constitutive equations of the fluid. In practice the constitutive equations cannot be fully stated without reference to the conservation laws. Two classes of model have been distinguished: mixture model and two-fluid models. In mixture models, the mixture is considered as a single fluid. Besides the usual friction factor and heat transfer correlations, a single constitutive law is necessary. In diffusion models, the mixture equation of state is replaced by the phasic equations of state and by three consitutive laws, for phase change mass transfer, drift velocity and thermal non-equilibrium respectively. In the two-fluid models, the two phases are considered separately; two phasic equations of state, two friction factor correlations, two heat transfer correlations and four constitutive laws are included [fr

  15. Fluidelastic instability analysis of steam generator U-tubes at antivibration bar-inactive modes

    International Nuclear Information System (INIS)

    Lee, S.K.; Jo, J.C.

    1995-01-01

    This paper presents the results of thermal-hydraulic and fluidelastic U-tube instability analyses performed for the vertical type pressurized water reactor (PWR) steam generator model being employed at Kori units 2, 3 and 4, and Yonggwang units 1 and 2 in Korea. The thermal-hydraulic analysis for providing the detailed three-dimensional two-phase flow field in the secondary side of the steam generator was accomplished using the ATHOS3 steam generator thermal-hydraulic analysis code. The UTVA2 code designed for calculating both the free vibration responses and fluidelastic stability ratio of a specific U-tube under consideration was used to assess the potential for fluidelastic instability of the steam generator U-tubes at various conditions of antivibration bar (AVB)-inactive modes. The results of the fluidelastic instability analysis were discussed in comparison with those obtained for the steam generator U-tubes at AVB-active mode

  16. Film boiling from spheres in single- and two-phase flow

    International Nuclear Information System (INIS)

    Liu, C.; Theofanous, T.G.; Yuen, W.W.

    1992-01-01

    Experimental data on film boiling heat transfer from single, inductively heated, spheres in single- and two-phase flow (saturated water and steam, respectively) are presented. In the single-phase-flow experiments water velocities ranged from 0.1 to 2.0 m/s; in the two-phase-flow experiments superficial water and steam velocities covered 0.1 to 0.6 m/s and 4 to 10 m/s, respectively. All experiments were run at atmospheric pressure and with sphere temperatures from 900C down to quenching. Limited interpretations of the single-phase- flow data are possible, but the two-phase-flow data are new and unique

  17. 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

  18. The Condensation effect on the two-phase flow stability

    International Nuclear Information System (INIS)

    Abdou Mohamed, Hesham Nagah

    2005-01-01

    considering riser condensation and of correcting the localized friction due to the presence of the two-phase mixture in the two-phase region.These effects are more important for high heating power and high inlet subcooling. CAREM 25 nuclear power reactor is investigated to get the stability boundary map. The flow instability regions are appeared at low and high core power. In the low heat flux range, the trends of the thermal equilibrium - equal velocity (homogeneous) model and the thermal non equilibrium - non equal velocity model are the same because the steam quality is small.In the high heat flux range, for the subcooled boiling number and the phase change number, the marginal stability boundaries are crossed in a point, determining tow different regions, of high and low inlet subcooling.For the first region, the steam quality calculation of the first model is greater and has the effect of stabilizing the system more than the second one.For the second region, the two-phase region length calculation of the first model is smaller and has the effect of stabilizing the system less than the second one. In general, the model predicts a more stable system with an increase in inlet restriction or riser condensation or system pressure or a decrease in exit restriction [es

  19. Comparative study of deterioration procedure in chemical-leavened steamed bread dough under frozen storage and freeze/thaw condition.

    Science.gov (United States)

    Wang, Pei; Yang, Runqiang; Gu, Zhenxin; Xu, Xueming; Jin, Zhengyu

    2017-08-15

    Successive freeze/thaw (FT) cycle was a widely used empirical approach to shorten the experimental period since it could accelerate frozen dough deterioration compared with frozen storage (FS). In order to compare the effect of FS and FT cycle on deterioration procedure of chemical-leavened steamed bread dough, kinetic studies of bread quality indices were performed and the relationships between bread quality and dough components were further established. Results showed that degradation of steamed bread loaf volume and firmness followed first-order kinetics during FS and zero-order kinetics during FT, respectively. Glutenin macropolymers (GMP) depolymerization and dough weight loss occurred steadily throughout FS and FT. Significant enhancement of damaged starch and crystallinity were observed at the later FS period and FT cycle. Multiple regression study led to the conclusion that dough weight loss contributed the most to the reduced bread loaf volume under FS whereas GMP depolymerization dominated under FT condition. Copyright © 2017 Elsevier Ltd. All rights reserved.

  20. A theoretical model for measuring mass flowrate and quality of two phase flow by the noise of throttling set

    International Nuclear Information System (INIS)

    Tong Yunxian; Wang Wenran

    1992-03-01

    The mass flowrate and steam quality measuring of two phase flowrate is an essential issue in the tests of loss-of-coolant accident (LOCA). The spatial stochastic distribution of phase concentration would cause a differential pressure noise when two phase flow is crossing a throttling set. Under the assumption of that the variance of disperse phase concentration is proportional to its mean phase concentration and by using the separated flow model of two phase flow, it has demonstrated that the variance of noise of differential pressure square root is approximately proportional to the flowrate of disperse phase. Thus, a theoretical model for measuring mass flowrate and quality of two phase flow by noise measurement is developed. It indicates that there is a possibility to measure two phase flowrate and steam quality by using the simple theoretical model and a single throttling set

  1. Gulping phenomena in transient countercurrent two-phase flow

    International Nuclear Information System (INIS)

    Tehrani, Ali A.K.

    2001-04-01

    Apart from previous work on countercurrent gas-liquid flow, transient tank drainage through horizontal off-take pipes is described, including experimental procedure, flow pattern on observations and countercurrent flow limitation results. A separate chapter is devoted to countercurrent two-phase flow in a pressurised water reactor hot-leg scaled model. Results concerning low head flooding, high head and loss of bowl flooding, transient draining of the steam generator and pressure variation and bubble detachment are presented. The following subjects are covered as well: draining of sealed tanks of vertical pipes, unsteady draining of closed vessel via vertical tube, unsteady filling of a closed vessel via vertical tube from a constant head reservoir. Practical significance of the results obtained is discussed

  2. Development of nuclear thermal hydraulic verification tests and evaluation technology - Development of the ultrasonic method for two-phase mixture level measurement in nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Kim, Sang Jae; Kim, Hyung Tae; Moon, Young Min [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    2000-04-01

    An ultrasonic method is developed for the measurement of the two-phase mixture level in the reactor vessel or steam generator. The ultrasonic method is selected among the several non-nuclear two-phase mixture level measurement methods through two steps of selection procedure. A commercial ultrasonic level measurement method is modified for application into the high temperature, pressure, and other conditions. The calculation method of the ultrasonic velocity is modified to consider the medium as the homogeneous mixture of air and steam, and to be applied into the high temperature and pressure conditions. The cross-correlation technique is adopted as a detection method to reduced the effects of the attenuation and the diffused reflection caused by surface fluctuation. The waveguides are developed to reduce the loss of echo and to remove the effects of obstructs. The present experimental study shows that the developed ultrasonic method measures the two-phase mixture level more accurately than the conventional methods do. 21 refs., 60 figs., 13 tabs. (Author)

  3. Preliminary applications of the new Neptune two-phase CFD solver to pressurized thermal shock investigations

    International Nuclear Information System (INIS)

    Boucker, M.; Laviaville, J.; Martin, A.; Bechaud, C.; Bestion, D.; Coste, P.

    2004-01-01

    The objective of this communication is to present some preliminary applications to pressurized thermal shock (PTS) investigations of the CFD (Computational Fluid Dynamics) two-phase flow solver of the new NEPTUNE thermal-hydraulics platform. In the framework of plant life extension, the Reactor Pressure Vessel (RPV) integrity is a major concern, and an important part of RPV integrity assessment is related to PTS analysis. In the case where the cold legs are partially filled with steam, it becomes a two-phase problem and new important effects occur, such as condensation due to the Emergency Core Cooling (ECC) injections of sub-cooled water. Thus, an advanced prediction of RPV thermal loading during these transients requires sophisticated two-phase, local scale, 3-dimensional codes. In that purpose, a program has been set up to extend the capabilities of the NEPTUNE two-phase CFD solver. A simple set of turbulence and condensation model for free surface steam-water flow has been tested in simulation of an ECC high pressure injection representing facility, using a full 3-dimensional mesh and the new NEPTUNE solver. Encouraging results have been obtained but it should be noticed that several sources of error can compensate for one another. Nevertheless, the computation presented here allows to be reasonable confident in the use of two-phase CFD in order to carry out refined analysis of two-phase PTS scenarios within the next years

  4. Ruthenium release modelling in air and steam atmospheres under severe accident conditions using the MAAP4 code

    International Nuclear Information System (INIS)

    Beuzet, Emilie; Lamy, Jean-Sylvestre; Perron, Hadrien; Simoni, Eric; Ducros, Gérard

    2012-01-01

    Highlights: ► We developed a new modelling of fuel oxidation and ruthenium release in the EDF version of the MAAP4 code. ► We validated this model against some VERCORS experiments. ► Ruthenium release prediction quantitatively and qualitatively well reproduced under air and steam atmospheres. - Abstract: In a nuclear power plant (NPP), a severe accident is a low probability sequence that can lead to core fusion and fission product (FP) release to the environment (source term). For instance during a loss-of-coolant accident, water vaporization and core uncovery can occur due to decay heat. These phenomena enhance core degradation and, subsequently, molten materials can relocate to the lower head of the vessel. Heat exchange between the debris and the vessel may cause its rupture and air ingress. After lower head failure, steam and air entering in the vessel can lead to degradation and oxidation of materials that are still intact in the core. Indeed, Zircaloy-4 cladding oxidation is very exothermic and fuel interaction with the cladding material can decrease its melting temperature by several hundred of Kelvin. FP release can thus be increased, noticeably that of ruthenium under oxidizing conditions. Ruthenium is of particular interest because of its high radio-toxicity due to 103 Ru and 106 Ru isotopes and its ability to form highly volatile compounds, even at room temperature, such as gaseous ruthenium tetra-oxide (RuO 4 ). It is consequently of great need to understand phenomena governing steam and air oxidation of the fuel and ruthenium release as prerequisites for the source term issues. A review of existing data on these phenomena shows relatively good understanding. In terms of oxygen affinity, the fuel is oxidized before ruthenium, from UO 2 to UO 2+x . Its oxidation is a rate-controlling surface exchange reaction with the atmosphere, so that the stoichiometric deviation and oxygen partial pressure increase. High temperatures combined with the presence

  5. Mechanistic multidimensional analysis of horizontal two-phase flows

    International Nuclear Information System (INIS)

    Tselishcheva, Elena A.; Antal, Steven P.; Podowski, Michael Z.

    2010-01-01

    The purpose of this paper is to discuss the results of analysis of two-phase flow in horizontal tubes. Two flow situations have been considered: gas/liquid flow in a long straight pipe, and similar flow conditions in a pipe with 90 deg. elbow. The theoretical approach utilizes a multifield modeling concept. A complete three-dimensional two-phase flow model has been implemented in a state-of-the-art computational multiphase fluid dynamics (CMFD) computer code, NPHASE. The overall model has been tested parametrically. Also, the results of NPHASE simulations have been compared against experimental data for a pipe with 90 deg. elbow.

  6. Experimental investigation two phase flow in direct methanol fuel cells

    International Nuclear Information System (INIS)

    Mat, M. D.; Kaplan, Y.; Celik, S.; Oeztural, A.

    2007-01-01

    Direct methanol fuel cells (DMFC) have received many attentions specifically for portable electronic applications since it utilize methanol which is in liquid form in atmospheric condition and high energy density of the methanol. Thus it eliminates the storage problem of hydrogen. It also eliminates humidification requirement of polymeric membrane which is a problem in PEM fuel cells. Some electronic companies introduced DMFC prototypes for portable electronic applications. Presence of carbon dioxide gases due to electrochemical reactions in anode makes the problem a two phase problem. A two phase flow may occur at cathode specifically at high current densities due to the excess water. Presence of gas phase in anode region and liquid phase in cathode region prevents diffusion of fuel and oxygen to the reaction sites thus reduces the performance of the system. Uncontrolled pressure buildup in anode region increases methanol crossover through membrane and adversely effect the performance. Two phase flow in both anode and cathode region is very effective in the performance of DMYC system and a detailed understanding of two phase flow for high performance DMFC systems. Although there are many theoretical and experimental studies available on the DMFC systems in the literature, only few studies consider problem as a two-phase flow problem. In this study, an experimental set up is developed and species distributions on system are measured with a gas chromatograph. System performance characteristics (V-I curves) is measured depending on the process parameters (temperature, fuel ad oxidant flow rates, methanol concentration etc)

  7. Microstructural studies on steam oxidised Zr-2.5%Nb pressure tube under simulated LOCA condition

    International Nuclear Information System (INIS)

    Banerjee, Suparna; Sawarn, Tapan K.; Pandit, K.M.; Anantharaman, S.; Srivastava, D.; Sah, D.N.

    2013-03-01

    Study of the microstructural evolution of Zr-2.5%Nb pressure tube material of Indian Pressurized Heavy Water Reactors (PHWRs) due to steam oxidation at high temperature (in the range 500-1050°C) was carried out on pressure tube coupons. Hydrogen pick up was less than 55 ppm in the samples oxidized at temperatures up to 850°C but high (250-400 ppm) in the samples oxidized in the β phase region (900°C and above). The microstructure of the samples oxidized above the α-Zr/β-Zr transition temperature showed from the surface inwards sequentially the presence of an oxide layer, an underlying oxygen stabilized α-Zr layer and a prior β-Zr phase containing hydride precipitates. An increase in the hardness was observed near the oxide-metal interface in the coupons oxidized above 900°C, due to formation of oxygen stabilized α-Zr layer. Higher hardness was also observed in the base metal in the samples oxidized at 1000 and 1050°C (author)

  8. Cold start-up condition model for heat recovery steam generators

    International Nuclear Information System (INIS)

    Sindareh-Esfahani, Peyman; Habibi-Siyahposh, Ehsan; Saffar-Avval, Majid; Ghaffari, Ali; Bakhtiari-Nejad, Firooz

    2014-01-01

    A dynamic modeling of Heat Recovery Steam Generator (HRSG) during cold start-up operation in Combined Cycle Power Plant (CCPP) is introduced. In order to characterize the essential dynamic behavior of the HRSG during cold start-up; Dynamic equations of all HRSG's components are developed based on energy and mass balances. To describe precisely the operation of HRSG; a method based on nonlinear estimated functions for thermodynamic properties is applied to estimate the model parameters. Model parameters are evaluated by a designed algorithm based on Genetic Algorithm (GA). A wide set of experimental data is used to validate HRSG model during cold start-up operation. The simulation results show the reliability and validity of the developed model for cold start-up operation. - Highlights: •Presenting a mathematical model for HRSGs cold start-up based on energy and mass balances. •A designed parameter identification algorithm based on GA is presented. •Application of experimental data in order to model and validate simulation results

  9. Two-phase flow modeling in the rod bundle subchannel analysis

    International Nuclear Information System (INIS)

    Hisashi, Ninokata

    2006-01-01

    the NASCA code capabilities for BT is described. There a combination of experimental and computational fluid dynamics approaches is undertaken to construct a two-phase fluid dynamics database. The experimental approach consists of 1) high-resolution air-water tests performed under the room-temperature and atmospheric pressure conditions for the inter-subchannel exchanges, three-dimensional behaviors of liquid films, and spacer effects; and 2) integral steam-water tests performed at high-temperature and at higher pressure. In the integral tests, state-of the- arts of multi-phase flow measurement technologies are applied in order to obtain local and instantaneous data that reveal underlying detailed physical processes including high resolution void distributions inside a 4 x 4 bundle, liquid film thickness and two-phase flow regime. The analytical approach consists of computational multi-phase fluid dynamics (CMFD) applicable to two-phase flows. A physical interpretation of the equilibrium two-phase flow redistribution inside a rod bundle is discussed that is considered to closely be related to the void drift phenomena. Identification of interactions among dominant factors is a main objective of the integral test and acquired data will be utilized in verifying the improved subchannel code. Construction of a complete set of two-phase fluid dynamics database will be made by supplementing missing data regions with the aid of numerical analyses. Dependency on important state variables is extracted from the database and prototype constitutive equations are going to be proposed in the final stage of the project. (author)

  10. Two-phase flow modeling in the rod bundle subchannel analysis

    International Nuclear Information System (INIS)

    Hisashi, Ninokata

    2004-01-01

    methodology adopted to improve the NASCA code capabilities for BT is described. There a combination of experimental and computational fluid dynamics approaches is undertaken to construct a two-phase fluid dynamics database. The experimental approach consists of 1) high-resolution air-water tests performed under the room-temperature and atmospheric pressure conditions for the inter-subchannel exchanges, three-dimensional behaviors of liquid films, and spacer effects; and 2) integral steam-water tests performed at high-temperature and at higher pressure. In the integral tests, state-of-the- arts of multi-phase flow measurement technologies are applied in order to obtain local and instantaneous data that reveal underlying detailed physical processes including high resolution void distributions inside a 4 x 4 bundle, liquid film thickness and two-phase flow regime. The analytical approach consists of computational multi-phase fluid dynamics (CMFD) applicable to two-phase flows. A physical interpretation of the equilibrium two-phase flow redistribution inside a rod bundle is discussed that is considered to closely be related to the void drift phenomena. Identification of interactions among dominant factors is a main objective of the integral test and acquired data will be utilized in verifying the improved subchannel code. Construction of a complete set of two-phase fluid dynamics database will be made by supplementing missing data regions with the aid of numerical analyses. Dependency on important state variables is extracted from the database and prototype constitutive equations are going to be proposed in the final stage of the project. (author)

  11. Fluid-Elastic Instability of U-Tube Bundle in Air-Water Two-Phase Flow

    International Nuclear Information System (INIS)

    Chu, In Cheol; Lee, Chang Hee; Yun, Young Jung; Chung, Heung June

    2007-03-01

    Using steam generator U-tube flow-induced vibration test facility, the flow-induced vibration characteristics of U-tube in row 34-44 and line 71-77 were investigated. Air and water at room temperature and near atmospheric pressure were used as working fluids. In the present experiments, followings were evaluated under two-phase cross-flow condition: the fundamental vibration responses and the critical gap velocity for a fluid-elastic instability of U-tubes, the damping ratio and hydrodynamic mass of U-tubes. In addition, the fluid-elastic instability factor, K, was preliminary assessed using Connors' relation. In the case of the U-tubes which are not supported by partial egg-crate in OPR100 steam generator, it has been found that the vibration displacement of those U-tubes are highly possible to exceed the design limit even by a turbulent excitation mechanism. The damping ratio of U-tubes measured in the present experiments was significantly higher than the OPR1000 steam generator design value. The fluid-elastic instability factor of U-tube bundle obtained in the present experiments were preliminary evaluated to be mostly in the range of 6.5-10.5

  12. Two phase cooling for superconducting magnets

    International Nuclear Information System (INIS)

    Eberhard, P.H.; Gibson, G.A.; Green, M.A.; Ross, R.R.; Smits, R.G.

    1986-01-01

    Comments on the use of two phase helium in a closed circuit tubular cooling system and some results obtained with the TPC superconducting magnet are given. Theoretical arguments and experimental evidence are given against a previously suggested method to determine helium two phase flow regimes. Two methods to reduce pressure in the magnet cooling tubes during quenches are discussed; 1) lowering the density of helium in the magnet cooling tubes and 2) proper location of pressure relief valves. Some techniques used to protect the refrigerator from too much cold return gas are also mentioned

  13. Two phase cooling for superconducting magnets

    International Nuclear Information System (INIS)

    Eberhard, P.H.; Gibson, G.A.; Green, M.A.; Ross, R.R.; Smits, R.G.; Taylor, J.D.; Watt, R.D.

    1986-01-01

    Comments on the use of two phase helium in a closed circuit tubular cooling system and some results obtained with the TPC superconducting magnet are given. Theoretical arguments and experimental evidence are given against a previously suggested method to determine helium two phase flow regimes. Two methods to reduce pressure in the magnet cooling tubes during quenches are discussed; (1) lowering the density of helium in the magnet cooling tubes and (2) proper location of pressure relief valves. Some techniques used to protect the refrigerator from too much cold return gas are also mentioned. 10 refs., 1 fig., 5 tabs

  14. Dynamic simulation of steam generator failures

    Energy Technology Data Exchange (ETDEWEB)

    Meister, G [Institut fuer Nukleare Sicherheitsforschung, Kernforschungsanlage Juelich GmbH, Juelich (Germany)

    1988-07-01

    A computer program will be described which is capable to simulate severe transients in a gas heated steam generator. Such transients may arise in the safety analysis of accidents resulting from failures in the heat removal system of an HTGR power plant. Important failure modes which have to be considered are ruptures of one or more steam generator tubes leading to water or steam ejection into the primary system or anomalous operating conditions which my cause damage due to excessive thermal stress. Examples are the complete dryout as a consequence of feedwater interrupt in connection with continuing gas heating and the reflooding of the secondary channel with cold feedwater after dryout. The steam generator program which is capable to simulate accidents of this type is written as a module which can be implemented into a program system fur the simulation of the total heat rejection system. It based on an advanced mathematical model for the two phase flow taking deviations from thermal equilibrium into account. Mass, energy and momentum balances for the primary and secondary fluid and the heat diffusion equations for the heat exchanging wall form a system of coupled differential equations which is solved numerically by an algorithm which is stiffly stable and suppresses effectively oscillations of numerical origin. Results of the simulation of transients of the type mentioned above will be presented and discussed. (author)

  15. Dynamic simulation of steam generator failures

    International Nuclear Information System (INIS)

    Meister, G.

    1988-01-01

    A computer program will be described which is capable to simulate severe transients in a gas heated steam generator. Such transients may arise in the safety analysis of accidents resulting from failures in the heat removal system of an HTGR power plant. Important failure modes which have to be considered are ruptures of one or more steam generator tubes leading to water or steam ejection into the primary system or anomalous operating conditions which my cause damage due to excessive thermal stress. Examples are the complete dryout as a consequence of feedwater interrupt in connection with continuing gas heating and the reflooding of the secondary channel with cold feedwater after dryout. The steam generator program which is capable to simulate accidents of this type is written as a module which can be implemented into a program system fur the simulation of the total heat rejection system. It based on an advanced mathematical model for the two phase flow taking deviations from thermal equilibrium into account. Mass, energy and momentum balances for the primary and secondary fluid and the heat diffusion equations for the heat exchanging wall form a system of coupled differential equations which is solved numerically by an algorithm which is stiffly stable and suppresses effectively oscillations of numerical origin. Results of the simulation of transients of the type mentioned above will be presented and discussed. (author)

  16. Two phase transitions in Nuclear Physics

    International Nuclear Information System (INIS)

    Bes, D.R.

    1985-01-01

    The status of the art of the problem associated with two phase transitions in the nuclear matter, viz.: the disappearance of the nuclear superfluiditiy with the raising of the rotation velocity and the appearance of an octupolar deformation in the actinide zone, is presented. (L.C.) [pt

  17. Two-phase flow in fractured rock

    International Nuclear Information System (INIS)

    Davies, P.; Long, J.; Zuidema, P.

    1993-11-01

    This report gives the results of a three-day workshop on two-phase flow in fractured rock. The workshop focused on two-phase flow processes that are important in geologic disposal of nuclear waste as experienced in a variety of repository settings. The goals and objectives of the workshop were threefold: exchange information; describe the current state of understanding; and identify research needs. The participants were divided into four subgroups. Each group was asked to address a series of two-phase flow processes. The following groups were defined to address these processes: basic flow processes; fracture/matrix interactions; complex flow processes; and coupled processes. For each process, the groups were asked to address these four issues: (1) describe the two-phase flow processes that are important with respect to repository performance; (2) describe how this process relates to the specific driving programmatic issues given above for nuclear waste storage; (3) evaluate the state of understanding for these processes; and (4) suggest additional research to address poorly understood processes relevant to repository performance. The reports from each of the four working groups are given here

  18. Numerical methods on flow instabilities in steam generator

    International Nuclear Information System (INIS)

    Yoshikawa, Ryuji; Hamada, Hirotsugu; Ohshima, Hiroyuki; Yanagisawa, Hideki

    2008-06-01

    The phenomenon of two-phase flow instability is important for the design and operation of many industrial systems and equipment, such as steam generators. The designer's job is to predict the threshold of flow instability in order to design around it or compensate for it. So it is essential to understand the physical phenomena governing such instability and to develop computational tools to model the dynamics of boiling systems. In Japan Atomic Energy Agency, investigations on heat transfer characteristics of steam generator are being performed for the development of Sodium-cooled Fast Breeder Reactor. As one part of the research work, the evaluations of two-phase flow instability in the steam generator are being carried out experimentally and numerically. In this report, the numerical methods were studied for two-phase flow instability analysis in steam generator. For numerical simulation purpose, the special algorithm to calculate inlet flow rate iteratively with inlet pressure and outlet pressure as boundary conditions for the density-wave instability analysis was established. There was no need to solve property derivatives and large matrices, so the spurious numerical instabilities caused by discontinuous property derivatives at boiling boundaries were avoided. Large time-step was possible. The flow instability in single heat transfer tube was successfully simulated with homogeneous equilibrium model by using the present algorithm. Then the drift-flux model including the effects of subcooled boiling and two phase slip was adopted to improve the accuracy. The computer code was developed after selecting the correlations of drift velocity and distribution parameter. The capability of drift flux model together with the present algorithm for simulating density-wave instability in single tube was confirmed. (author)

  19. Analysis of two-phase flow and boiling heat transfer in inclined channel of core-catcher

    International Nuclear Information System (INIS)

    Tahara, M.; Suzuki, Y.; Abe, N.; Kurita, T.; Hamazaki, R.; Kojima, Y.

    2008-01-01

    Passive Corium Cooling System (CCS) provides a function of ex-vessel debris cooling and molten core stabilization during a severe accident. CCS features inclined cooling channels arranged axi-symmetrically below the core-catcher basin. In order to estimate the coolability of the inclined cooling channel, it is indispensable to identify the flow pattern of the two-phase flow in the cooling channel. Several former studies for the two-phase flow pattern in the inclined channel are referred. Taitel and Dukler (1976) developed a prediction method of the flow pattern transition in horizontal and near horizontal tubes. Barnea et al. (1980) showed the flow pattern map of upward flow with 10 degrees inclination. Sakaguti et al. (1996) observed the two-phase flow patterns in the horizontal pipe connected with slightly upward pipe, in which the flow pattern in the pipe with a bending part was expressed by the combination of a basic flow pattern and some auxiliary flow patterns. Then we investigated these studies In order to identify the flow patterns observed in the inclined cooling channel of CCS. Furthermore we experimentally observed the flow patterns in the inclined cooling channel with various inlet conditions. As a result of the investigation and observation, typical flow patterns in the inclined cooling channel were identified. Two typical flow patterns were observed depending on the steam flow rate, one of which is 'elongated bubble 'flow, and the other is 'churn with collapsing backward and upward slug 'flow The flow and heat transfer in the inclined channel of CCS is analyzed by using a two-phase analysis code employing two-fluid model in which the constitutive equations for the two-phase flow in inclined channels are incorporated. That is, drift flux parameter for each of the elongated bubble flow, and the churn with collapsing backward and upward slug flow are incorporated to the two-phase analysis code, which are based on the rising velocity of the long bubble in

  20. Thermal-hydraulics in recirculating steam generators

    International Nuclear Information System (INIS)

    Carver, M.B.; Carlucci, L.N.; Inch, W.W.R.

    1981-04-01

    This manual describes the THIRST code and its use in computing three-dimensional two-phase flow and heat transfer in a steam generator under steady state operation. The manual is intended primarily to facilitate the application of the code to the analysis of steam generators typical of CANDU nuclear stations. Application to other steam generator designs is also discussed. Details of the assumptions used to formulate the model and to implement the numerical solution are also included

  1. Thermo-hydraulic stability study of a steam generator

    International Nuclear Information System (INIS)

    Magni, M C; Marcel, C P; Delmastro, D F

    2012-01-01

    In this work a mathematical model developed to investigate the thermalhydraulic stability of a helically coiled steam generator is presented. Such a steam generator is prone to experiment density wave oscillations. The model is therefore used to analyze the stability of the CAREM-25 reactor steam generators. The model is linear, numerically non-diffusive and nodal. In addition, it is able to represent non-uniform heat transfer fluxes between the primary and secondary coolant circuits. By using this model the marginal stability condition is found by varying the inlet friction coefficient for different conditions. The results are then compared with those obtained with a different model for which a simple uniform heat flux profiled is assumed. It is found that with such simplification the density waves instability mechanism is overestimated in a wide range of operating powers. For very low powers, in the contrary, the so-called uniform model underestimates the stabilizing inlet friction and therefore it gives non-conservative results. With the use of the more realistic non-uniform power profile model, it was possible to determine that, for a CAREM-25 steam generator, the most stable conditions is found at 60MW when the reactor operates at nominal pressure. Moreover, it is found that at high power levels the stability performance is dominated by the two-phase friction component while at low power levels the friction component originated in the over heated steam region prevail (author)

  2. Peculiarities of cyclic deformation and fracture of heat-resistant steel 10GN2MFA under conditions typical for the steam generator PGV-1000 collector material

    International Nuclear Information System (INIS)

    Giginyak, F.F.

    1998-01-01

    In the present paper, the results are discussed concerning investigations into the regularities of deformation and fracture of steel, 10GN2MFA under conditions close to those of actual operation when used in collectors of PGV-1000-type steam generators, which are in service with WWER-1000-type reactors of nuclear power plants (NPP). (author)

  3. Heat and mass transfer and hydrodynamics in two-phase flows in nuclear power plants

    International Nuclear Information System (INIS)

    Styrikovich, M.A.; Polonskii, V.S.; Tsiklauri, G.V.

    1986-01-01

    This book examines nuclear power plant equipment from the point of view of heat and mass transfer and the behavior of impurities contained in water and in steam, with reference to real water regimes of nuclear power plants. The transfer processes of equipment are considered. Heat and mass transfer are analyzed in the pre-crisis regions of steam-generating passages with non-permeable surfaces, and in capillary-porous structures. Attention is given to forced convection boiling crises and top post-DNB heat transfer. Data on two-phase hydrodynamics in straight and curved channels are correlated and safety aspects of nuclear power plants are discussed

  4. Raman scattering temperature measurements for water vapor in nonequilibrium dispersed two-phase flow

    International Nuclear Information System (INIS)

    Anastasia, C.M.; Neti, S.; Smith, W.R.; Chen, J.C.

    1982-09-01

    The objective of this investigation was to determine the feasibility of using Raman scattering as a nonintrusive technique to measure vapor temperatures in dispersed two-phase flow. The Raman system developed for this investigation is described, including alignment of optics and optimization of the photodetector for photon pulse counting. Experimentally obtained Raman spectra are presented for the following single- and two-phase samples: liquid water, atmospheric nitrogen, superheated steam, nitrogen and water droplets in a high void fraction air/water mist, and superheated water vapor in nonequilibrium dispersed flow

  5. Small break critical discharge: The roles of vapor and liquid entrainment in a stratified two-phase region upstream of the break

    International Nuclear Information System (INIS)

    Schrock, V.E.; Revankar, S.T.; Mannheimer, R.; Wang, C.H.

    1986-12-01

    The main objective of this research program was to perform an experimental investigation on the phenomena of two-phase critical flow through small break from a horizontal pipe which contained a stratified two phase flow. Stagnation conditions investigated were saturated steam-water, and air-cold water at pressures ranging from 0.37 MPa to 1.07 MPa. The small breaks employed were cylindrical tubes of diameters 3.96 mm, 6.32 mm, and 10.1 mm with sharp-edged entrance. For breaks resulting from a small hole in a primary coolant pipe or in a small pipe, a sharp-edged orifice or a sharp-edged tube can be the approximation

  6. An introduction to two-phase flows

    International Nuclear Information System (INIS)

    Lemonnier, Herve

    2006-01-01

    This course aims at proposing the necessary background for a rational approach to two-phase flows which are notably present in numerous industrial devices and equipment designed to perform energy transfer or mass transfer. The first part proposes a phenomenological approach to main two-phase flow structures and presents their governing variables. The second part presents some proven measurement techniques. The third part focuses on modelling. It recalls the equation elaboration techniques which are based on basic principles of mechanics and thermodynamics and on the application of different averaging operators to these principles. Some useful models are then presented such as models of pressure loss in a duct. The last chapter addresses some fundamental elements of heat transfers in ebullition and condensation

  7. Apparatus for monitoring two-phase flow

    Science.gov (United States)

    Sheppard, John D.; Tong, Long S.

    1977-03-01

    A method and apparatus for monitoring two-phase flow is provided that is particularly related to the monitoring of transient two-phase (liquid-vapor) flow rates such as may occur during a pressurized water reactor core blow-down. The present invention essentially comprises the use of flanged wire screens or similar devices, such as perforated plates, to produce certain desirable effects in the flow regime for monitoring purposes. One desirable effect is a measurable and reproducible pressure drop across the screen. The pressure drop can be characterized for various known flow rates and then used to monitor nonhomogeneous flow regimes. Another useful effect of the use of screens or plates in nonhomogeneous flow is that such apparatus tends to create a uniformly dispersed flow regime in the immediate downstream vicinity. This is a desirable effect because it usually increases the accuracy of flow rate measurements determined by conventional methods.

  8. Geometrical automata for two phase flow simulation

    International Nuclear Information System (INIS)

    Herrero, V.; Guido-Lavalle, G.; Clausse, A.

    1996-01-01

    An automaton is an entity defined by a mathematical state which changes following iterative rules representing the interaction with the neighborhood. A model of automata for two-phase flow simulation consisting in a field of disks which are allowed to change their radii and move in a plane is presented. The model is more general than the classical cellular automata in two respects: (1) the grid of cellular automata is dismissed in favor of a trajectory generator; and (2) the rules of interaction involve parameters intended to represent some of the most relevant variables governing the actual physical interactions between phases. Computational experiments show that the algorithm captures the essential physics underlying two-phase flow problems such as bubbly-slug pattern transition and void fraction development along tubes. A comparison with experimental data of void fraction profiles is presented, showing excellent agreement. (orig.)

  9. Apparatus for monitoring two-phase flow

    International Nuclear Information System (INIS)

    Sheppard, J.D.; Tong, L.S.

    1977-01-01

    A method and apparatus for monitoring two-phase flow is provided that is particularly related to the monitoring of transient two-phase (liquid-vapor) flow rates such as may occur during a pressurized water reactor core blow-down. The present invention essentially comprises the use of flanged wire screens or similar devices, such as perforated plates, to produce certain desirable effects in the flow regime for monitoring purposes. One desirable effect is a measurable and reproducible pressure drop across the screen. The pressure drop can be characterized for various known flow rates and then used to monitor nonhomogeneous flow regimes. Another useful effect of the use of screens or plates in nonhomogeneous flow is that such apparatus tends to create a uniformly dispersed flow regime in the immediate downstream vicinity. This is a desirable effect because it usually increases the accuracy of flow rate measurements determined by conventional methods. 3 claims, 9 figures

  10. Pumped two-phase heat transfer loop

    Science.gov (United States)

    Edelstein, Fred

    1988-01-01

    A pumped loop two-phase heat transfer system, operating at a nearly constant temperature throughout, includes several independently operating grooved capillary heat exchanger plates supplied with working fluid through independent flow modulation valves connected to a liquid supply line, a vapor line for collecting vapor from the heat exchangers, a condenser between the vapor and the liquid lines, and a fluid circulating pump between the condenser and the heat exchangers.

  11. Visualization of Two Phase Natural Convection Flow in a Vertical Pipe using the Sulfuric Acid - Copper Sulfate Electroplating System

    Energy Technology Data Exchange (ETDEWEB)

    Ohk, Seung-Min; Chae, Myeong-Seon; Chung, Bum-Jin [Kyung Hee Univ., Yongin (Korea, Republic of)

    2014-10-15

    The passive containment cooling system (PCCS) driven by natural forces convection gain draws research interests after Fukushima NPP accident. The PCCS was classified into three categories: Containment pressure suppression, Containment passive heat removal/pressure suppression systems and Passive containment spray. Among the types of containment passive heat removal/pressure suppression systems, the system composed of an internal heat exchanger and an external coolant tank is considered. In a severe accident condition, the heat from the containment atmosphere is transferred to the outer surface of the heat exchanger by the convection and condensation of the mixture of steam and gases. On the other hand, the heat is transferred to external pool by single phase or two phase natural convection inside of heat exchanger pipes. The study aimed at investigating the influence of the diameter (D) and height (H) of the heat exchanger pipes on the single phase and two phase natural convection heat transfer. As the initial stage of the study, the two phase natural convection flow inside a vertical pipe is visualized. In order to achieve the aim with ample test rig, a sulfuric acid - cooper sulfate electroplating system was employed based on the analogy between heat and mass transfer. The reduction of hydrogen ion at the cathode surface at high potential was used to simulate the boiling phenomena. This study tried to visualize the boiling heat transfer inside a vertical pipe using a cupric acid-copper sulfate (H{sub 2}SO{sub 4}-CuSO{sub 4}) electroplating system. This seems to be successful so far. However further study has to be done to compare the result with real two phase flow situation. The surface tension and surface characteristics are to be tuned to simulate the real situation.

  12. A new correlation for two-phase critical discharge coefficient

    International Nuclear Information System (INIS)

    Park, Jong Woon; Chun, Moon Hyun

    1989-01-01

    A new simple correlation for subcooled and two-phase critical flow discharge coefficient has been developed by stepwise regression technique. The new discharge coefficient has three independent variables and they are length to hydraulic diameter ratio, degree of subcooling, and stagnation temperature. The new discharge coefficient is applied as a multiplier to homogeneous equilibrium model and Abauf's single phase critical mass flux calculation equation. This method has been tested for its accuracy by comparing with experimental data. Results of the comparison show that the agreement between the predictions with new correlation and the experimental data is good for pipes and nozzles with vertical upward flow for subcooled upstream condition and nozzles with horizontal configuration for two-phase upstream condition

  13. Visualization of large waves in churn and annular two-phase flow

    International Nuclear Information System (INIS)

    Dasgupta, Arnab; Chandraker, D.K.; Nayak, A.K.; Vijayan, P.K.; Kshirasagar, S.; Reddy, B.R.; Walker, S.P.

    2015-01-01

    The study of churn and annular two-phase flow regimes is important for boiling systems like nuclear reactors, U-tube steam generators etc. In this paper, visualization studies on air-water churn and annular two-phase flow regimes are reported. Though there are differences between air-water and boiling steam water systems, the major flow-pattern characteristics are similar (if not same).The specific object of study is the large waves which exist in both churn and annular regimes. These waves are responsible for majority of the momentum and mass dispersion across the phases. The differentiating characteristics of these waves in the chum and annular flow regimes are reported. The visualization also leads to a more quantitative representation of the transition from churn to annular flow. A new interpretation of the criterion for onset of entrainment is also evolved from the studies. (author)

  14. Measurement of transient two-phase flow velocity using statistical signal analysis of impedance probe signals

    International Nuclear Information System (INIS)

    Leavell, W.H.; Mullens, J.A.

    1981-01-01

    A computational algorithm has been developed to measure transient, phase-interface velocity in two-phase, steam-water systems. The algorithm will be used to measure the transient velocity of steam-water mixture during simulated PWR reflood experiments. By utilizing signals produced by two, spatially separated impedance probes immersed in a two-phase mixture, the algorithm computes the average transit time of mixture fluctuations moving between the two probes. This transit time is computed by first, measuring the phase shift between the two probe signals after transformation to the frequency domain and then computing the phase shift slope by a weighted least-squares fitting technique. Our algorithm, which has been tested with both simulated and real data, is able to accurately track velocity transients as fast as 4 m/s/s

  15. Theoretical investigation of flow regime for boiling water two-phase flow in horizontal rectangular narrow channels

    International Nuclear Information System (INIS)

    Zhang Chunwei; Qiu Suizheng; Yan Mingyu; Wang Bulei; Nie Changhua

    2005-01-01

    The flow regime transition criteria for the boiling water two-phase flow in horizontal rectangular narrow channels (1 x 20 mm, 2 x 20 mm) were theoretically explored. The discernible flow patterns were bubble, intermittent slug, churn, annular and steam-water separation flow. By using two-fluid model, equations of conservation of momentum were established for the two-phase flow. New flow-regime criteria were obtained and agreed well with the experiment data. (authors)

  16. An experimental study of the heat transfer performance of a rectangular two-phase natural circulation loop

    International Nuclear Information System (INIS)

    Chen, K.S.; Chen, Y.Y.; Tsai, S.T.

    1990-01-01

    An experimental study is presented for the heat transfer performance of a rectangular, two-phase, natural-circulation loop with water-steam as the working fluid. Local temperature measurements of the core fluid and the wall were made, and the overall heat transfer coefficients of the evaporator, the condenser, and the loop system were obtained and correlated in terms of the fluid properties, heat flux conditions, and the liquid charge level. An overheat phenomenon at very low charge level was also observed. Result of a preliminary analysis shows that if the liquid charge level is below the fractional volume of the connecting tube between the condenser and the evaporator, an overheat phenomenon will occur

  17. Governing equations for a seriated continuum: an unequal velocity model for two-phase flow

    International Nuclear Information System (INIS)

    Solbrig, C.W.; Hughes, E.D.

    1975-05-01

    The description of the flow of two-phase fluids is important in many engineering devices. Unexpected transient conditions which occur in these devices cannot, in general, be treated with single-component momentum equations. Instead, the use of momentum equations for each phase is necessary in order to describe the varied transient situations which can occur. These transient conditions can include phases moving in the opposite directions, such as steam moving upward and liquid moving downward, as well as phases moving in the same direction. The derivation of continuity and momentum equations for each phase and an overall energy equation for the mixture are presented. Terms describing interphase forces are described. A seriated (series of) continuum is distinguished from an interpenetrating medium by the representation of interphase friction with velocity differences in the former and velocity gradients in the latter. The seriated continuum also considers imbedded stationary solid surfaces such as occur in nuclear reactor cores. These stationary surfaces are taken into account with source terms. Sufficient constitutive equations are presented to form a complete set of equations. Methods are presented to show that all these coefficients are determinable from microscopic models and well known experimental results. Comparison of the present deviation with previous work is also given. The equations derived here may also be employed in certain multiphase, multicomponent flow applications. (U.S.)

  18. Design Development of SMART ECC Water Asymmetric Two-phase choking test facility

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young Il; Cho, Seok; Ko, Yung Joo; Shin, Yong Cheol; Kwon, Tae Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    SMART pressurized water reactor type is different from the existing integral NSSS commercial pressurized water reactor system which is equipped with the main features. In addition RCS piping is removed and the feature of the SBLOCA is a major design break accident. The TASS / SMR code is analyzed SMART SBLOCA. In order to verify analysis code, SMART analysis for verification of conservatism is promoting using data for experiments with Integral Effect Test and Separate Effect. In this paper, the design feature of the SWAT (SMART ECC Water Asymmetric Two-phase choking test facility) is described. SWAT is linearly reduced to a 1/5 ratio while the geometrical shape is conserved. In major shape of SMART ECC injection performance test, distortions which caused by gravitational effects are minimized. Because both the emergency core cooling water injection nozzle height and the break nozzle height match the RCP Suction Nozzle height in test section of the main forms. The main part of the test section is SG-side upper down-comer. The boundary conditions are saturated steam and water flow condition and drain flow rate to control the collapsed water level in the down-comer

  19. Two-phase turbulent mixing and buoyancy drift in rod bundles

    International Nuclear Information System (INIS)

    Carlucci, L.N.; Hammouda, N.; Rowe, D.S.

    2004-01-01

    This paper describes the development of generalized relationships for single- and two-phase inter subchannel turbulent mixing in vertical and horizontal flows, and lateral buoyancy drift in horizontal flows. The relationships for turbulent mixing, together with a recommended one for void drift, have been implemented in a subchannel thermal hydraulics code, and assessed using a range of data on enthalpy migration in vertical steam-water lows under BWR and PWR diabatic conditions. The intent of this assessment as to optimize these relationships to give the best agreement with the enthalpy migration data for vertical flows. The optimized turbulent mixing relationships were then used as a basis to benchmark a proposed buoyancy rift model to give the best predictions of void and enthalpy migration data n horizontal flows typical of PHWR CANDU reactor operation under normal and off-normal conditions. Overall, the optimized turbulent mixing and buoyancy drift relationships have been found to predict the available data quite well, nd generally better and more consistently than currently used models. This is expected to result in more accurate calculations of subchannel distributions of phasic flows, and hence, in improved predictions of critical heat flux (CHF)

  20. Instrumentation for two-phase flow measurements in code verification experiments

    International Nuclear Information System (INIS)

    Fincke, J.R.; Anderson, J.L.; Arave, A.E.; Deason, V.A.; Lassahn, G.D.; Goodrich, L.D.; Colson, J.B.; Fickas, E.T.

    1981-01-01

    The development of instrumentation and techniques for the measurement of mass flow rate in two-phase flows conducted at the Idaho National Engineering Laboratory during the past year is briefly described. Instruments discussed are the modular drag-disc turbine transducer, the gamma densitometers, the ultrasonic densitometer, Pitot tubes, and full-flow drag screens. Steady state air-water and transient steam-water data are presented

  1. A study of critical two-phase flow models

    International Nuclear Information System (INIS)

    Siikonen, T.

    1982-01-01

    The existing computer codes use different boundary conditions in the calculation of critical two-phase flow. In the present study these boundary conditions are compared. It is shown that the boundary condition should be determined from the hydraulic model used in the computer code. The use of a correlation, which is not based on the hydraulic model used, leads often to bad results. Usually a good agreement with data is obtained in the calculation as far as the critical mass flux is concerned, but the agreement is not so good in the pressure profiles. The reason is suggested to be mainly in inadequate modeling of non-equilibrium effects. (orig.)

  2. Reynolds stress turbulence model applied to two-phase pressurized thermal shocks in nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Mérigoux, Nicolas, E-mail: nicolas.merigoux@edf.fr; Laviéville, Jérôme; Mimouni, Stéphane; Guingo, Mathieu; Baudry, Cyril

    2016-04-01

    Highlights: • NEPTUNE-CFD is used to model two-phase PTS. • k-ε model did produce some satisfactory results but also highlights some weaknesses. • A more advanced turbulence model has been developed, validated and applied for PTS. • Coupled with LIM, the first results confirmed the increased accuracy of the approach. - Abstract: Nuclear power plants are subjected to a variety of ageing mechanisms and, at the same time, exposed to potential pressurized thermal shock (PTS) – characterized by a rapid cooling of the internal Reactor Pressure Vessel (RPV) surface. In this context, NEPTUNE-CFD is used to model two-phase PTS and give an assessment on the structural integrity of the RPV. The first available choice was to use standard first order turbulence model (k-ε) to model high-Reynolds number flows encountered in Pressurized Water Reactor (PWR) primary circuits. In a first attempt, the use of k-ε model did produce some satisfactory results in terms of condensation rate and temperature field distribution on integral experiments, but also highlights some weaknesses in the way to model highly anisotropic turbulence. One way to improve the turbulence prediction – and consequently the temperature field distribution – is to opt for more advanced Reynolds Stress turbulence Model. After various verification and validation steps on separated effects cases – co-current air/steam-water stratified flows in rectangular channels, water jet impingements on water pool free surfaces – this Reynolds Stress turbulence Model (R{sub ij}-ε SSG) has been applied for the first time to thermal free surface flows under industrial conditions on COSI and TOPFLOW-PTS experiments. Coupled with the Large Interface Model, the first results confirmed the adequacy and increased accuracy of the approach in an industrial context.

  3. Pressure distribution over tube surfaces of tube bundle subjected to two phase cross flow

    International Nuclear Information System (INIS)

    Sim, Woo Gun

    2013-01-01

    Two phase vapor liquid flows exist in many shell and tube heat exchangers such as condensers, evaporators and nuclear steam generators. To understand the fluid dynamic forces acting on a structure subjected to a two phase flow, it is essential to obtain detailed information about the characteristics of a two phase flow. The characteristics of a two phase flow and the flow parameters were introduced, and then, an experiment was performed to evaluate the pressure loss in the tube bundles and the fluid dynamic force acting on the cylinder owing to the pressure distribution. A two phase flow was pre mixed at the entrance of the test section, and the experiments were undertaken using a normal triangular array of cylinders subjected to a two phase cross flow. The pressure loss along the flow direction in the tube bundles was measured to calculate the two phase friction multiplier, and the multiplier was compared with the analytical value. Furthermore, the circular distributions of the pressure on the cylinders were measured. Based on the distribution and the fundamental theory of two phase flow, the effects of the void fraction and mass flux per unit area on the pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure on the tube by a numerical method. It was found that for low mass fluxes, the measured two phase friction multipliers agree well with the analytical results, and good agreement for the effect of the void fraction on the drag coefficients, as calculated by the measured pressure distributions, is shown qualitatively, as compared to the existing experimental results

  4. Two-phase flow dynamics in ECC

    International Nuclear Information System (INIS)

    Albraaten, P.J.

    1981-07-01

    The present report summarizes the achievements within the project ''Two-phase Systems and ECC''. The results during 1978 - 1980 are accounted for in brief as they have been documented in earlier reports. The results during the first half of 1981 are accounted for in greater detail. They contain a new model for the Basset force and test runs with this model using the test code RISQUE. Furthermore, test runs have been performed with TRAC-PD2 MOD 1. This code was implemented on Edwards Pipe Blowdown experiment (a standard test case) and UC-Berkeley Reflooding experiment (a non-standard test case.) (Auth.)

  5. Two phase cooling for superconducting magnets

    International Nuclear Information System (INIS)

    Eberhard, P.H.; Gibson, G.A.; Green, M.A.; Ross, R.R.; Smits, R.G.; Taylor, J.D.; Watt, R.D.

    1985-08-01

    A closed circuit tubular cooling system for superconducting magnets offers advantages of limiting boiloff and containing high pressures during quenches. Proper location of automatic valves to lower pressures and protect the refrigerator in the event of quenches is described. Theoretical arguments and exprimental evidence are given against a previously suggested method to determine He two phase flow regimes. If loss of flow occurs due to some types of refrigeration failure and transfer lines have enough heat leak to warm up, quenches are induced when the flow is restored. Examples are taken from experience with the TPC magnet

  6. A theoretical critical heat flux model for low-pressure, low-mass-flux, and low-steam quality conditions

    International Nuclear Information System (INIS)

    Weihsiao Ho; Kuanchywan Tu; Baushei Pei; Chinjang Chang

    1993-01-01

    The critical heat flux (CHF) is the maximum heat flux just before a boiling crisis; its importance as a measurement of nuclear reactor power capability design as well as in the safety of reactors has been recognized. With emphasis on CHF behavior under subcooled and low-quality (i.e., 2 ·s), an improved model that uses the sublayer dry out theory has been developed. Based on experimental observations of CHF, the model assumes that CHF under such conditions is of the departure from nucleate boiling type. Based on the postulation that CHF is triggered by Helmholtz instability in the sublayer steam-liquid system, the model was developed by a simple energy balance of liquid sublayer evaporation as the vapor blanket tends to disturb the balance between the buoyancy force and the drag force exerted upon it. The model is compared with the well-known Biasi et al. correlation as well as the Atomic Energy of Canada Limited lookup table against 102 uniformly heated round tube CHF data and 34 nonuniformly heated round tube CHF data. The comparison shows that the model provides better accuracy and a reasonable agreement between the predicted values and experimental CHF data

  7. Development of evaluation method on flow-induced vibration and corrosion of components in two-phase flow by coupled analysis. 2. Evaluation of corrosive conditions in cooling systems of nuclear power plants

    International Nuclear Information System (INIS)

    Uchida, Shunsuke; Naitoh, Masanori; Okada, Hidetoshi; Uehara, Yasushi; Katono, Kenichi; Ichikawa, Ryoko; Hotta, Koji

    2007-01-01

    In order to establish safety and reliable plant operation, it is required to forecast future troubles on structural materials based on combined analyses of flow dynamics and corrosion and to mitigate them before they become serious problems on plant operation. The models are divided into two parts, which are as follows. 1. Prediction model of future trouble on structural materials based on evaluation of corrosive conditions. 2. Evaluation model of wall thinning caused by flow accelerated corrosion. (author)

  8. Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly

    International Nuclear Information System (INIS)

    Andrey Ioilev; Maskhud Samigulin; Vasily Ustinenko; Simon Lo; Adrian Tentner

    2005-01-01

    Full text of publication follows: The goal of this project is to develop an advanced Computational Fluid Dynamics (CFD) computer code (CFD-BWR) that allows the detailed analysis of the two-phase flow and heat transfer phenomena in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. This code will include more fundamental physical models than the current generation of sub-channel codes and advanced numerical algorithms for improved computational accuracy, robustness, and speed. It is highly desirable to understand the detailed two-phase flow phenomena inside a BWR fuel bundle. These phenomena include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for the analysis of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is still too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Recent progress in Computational Fluid Dynamics (CFD), coupled with the rapidly increasing computational power of massively parallel computers, shows promising potential for the fine-mesh, detailed simulation of fuel assembly two-phase flow phenomena. However, the phenomenological models available in the commercial CFD programs are not as advanced as those currently being used in the sub-channel codes used in the nuclear industry. In particular, there are no models currently available which are able to reliably predict the nature of the flow regimes, and use the appropriate sub-models for those flow regimes. The CFD-BWR code is being developed as a customized module built on the foundation of the commercial CFD Code STAR-CD which provides general two-phase flow modeling capabilities. The paper describes the model development strategy which has been adopted by the development team for the

  9. Optimization of wet lay-up conditions for steam generators hydrazine chemical treatment

    International Nuclear Information System (INIS)

    Long, A.; Organista, M.; Brun, C.; Combrade, P.

    2002-01-01

    Since a long time, hydrazine is used as a chemical agent to prevent corrosion of unalloyed steels. This is a conventional treatment widely used by nuclear power plant operators. But its application in SG lay-up at French nuclear power plants has, however, lead to some drawbacks. Effluent releases: Due to regulation relative to release of hydrazine and alkaline chemical compounds, some plant operators limit the concentrations of reagents to levels that could lead to insufficient protection of materials. Safety hazards associated with SG nitrogen blanketing: Prohibiting use of nitrogen blankets for SG wet lay-up due to associated safety hazards could likewise jeopardize corrosion protection at normally specified hydrazine levels. As the exact limits of hydrazine action against corrosion during SG lay-up are not well known, it is sometimes difficult to evaluate the risk associated to low dosage of N 2 H 4 . In order to answer to these problems, Framatome ANP (France) decided to carry out a test program aimed to determine the limit conditions for use of hydrazine in a wet lay-up environment. (authors)

  10. Two Phase Flow Simulation Using Cellular Automata

    International Nuclear Information System (INIS)

    Marcel, C.P.

    2002-01-01

    The classical mathematical treatment of two-phase flows is based on the average of the conservation equations for each phase.In this work, a complementary approach to the modeling of these systems based on statistical population balances of aut omata sets is presented.Automata are entities defined by mathematical states that change following iterative rules representing interactions with the neighborhood.A model of automata for two-phase flow simulation is presented.This model consists of fie lds of virtual spheres that change their volumes and move around a certain environment.The model is more general than the classical cellular automata in two respects: the grid of cellular automata is dismissed in favor of a trajectory generator, and the rules of interaction involve parameters representing the actual physical interactions between phases.Automata simulation was used to study unsolved two-phase flow problems involving high heat flux rates. One system described in this work consists of a vertical channel with saturated water at normal pressure heated from the lower surface.The heater causes water to boil and starts the bubble production.We used cellular automata to describe two-phase flows and the interaction with the heater.General rule s for such cellular automata representing bubbles moving in stagnant liquid were used, with special attention to correct modeling of different mechanisms of heat transfer.The results of the model were compared to previous experiments and correlations finding good agreement.One of the most important findings is the confirmation of Kutateladze's idea about a close relation between the start of critical heat flux and a change in the flow's topology.This was analyzed using a control volume located in the upper surface of the heater.A strong decrease in the interfacial surface just before the CHF start was encountered.The automata describe quite well some characteristic parameters such as the shape of the local void fraction in the

  11. Numerical simulation for two-phase jet problem

    International Nuclear Information System (INIS)

    Lee, W.H.; Shah, V.L.

    1981-01-01

    A computer program TWOP was developed for obtaining the numerical solutions of three-dimensional, transient, two-phase flow system with nonequilibrium and nonhomogeneous conditions. TWOP employs two-fluid model and a set of the conservation equations formulated by Harlow and Amsden along with their Implicit Multi-Field (IMF) numerical technique that allows all degrees of couplings between the two fields. We have further extended the procedure of Harlow and Amsden by incorporating the implicit couplings of phase transition and interfacial heat transfer terms in the energy equations. Numerical results of two tested problems are presented to demonstrate the capabilities of the TWOP code. The first problem is the separation of vapor and liquid, showing that the code can handle the computational difficulties such as liquid packing and sharp interface phenomena. The second problem is the high pressure two-phase jet impinged on vertical plate, demonstrating the important role of the interfacial mass and momentum exchange

  12. Study on flow instabilities in two-phase mixtures

    International Nuclear Information System (INIS)

    Ishii, M.

    1976-03-01

    Various mechanisms that can induce flow instabilities in two-phase flow systems are reviewed and their relative importance discussed. In view of their practical importance, the density-wave instabilities have been analyzed in detail based on the one-dimensional two-phase flow formulation. The dynamic response of the system to the inlet flow perturbations has been derived from the model; thus the characteristic equation that predicts the onset of instabilities has been obtained. The effects of various system parameters, such as the heat flux, subcooling, pressure, inlet velocity, inlet orificing, and exit orificing on the stability boundary have been analyzed. In addition to numerical solutions, some simple stability criteria under particular conditions have been obtained. Both results have been compared with various experimental data, and a satisfactory agreement has been demonstrated

  13. Mixed convection in a two-phase flow cooling loop

    International Nuclear Information System (INIS)

    Janssens-Maenhout, G.; Daubner, M.; Knebel, J.U.

    2002-03-01

    construction, can be directly applied to the design of accelerator driven systems (ADS). However, the transfer from the two-phase system water / steam to the system lead-bismuth / nitrogen still has to be done. Thus, the results of the POOLTHY Project are partly of direct relevance to the Programme on Partitioning and Transmutation and especially to the PDS-XADS Project. (orig.)

  14. Mixed convection in a two-phase flow cooling loop

    Energy Technology Data Exchange (ETDEWEB)

    Janssens-Maenhout, G.; Daubner, M.; Knebel, J.U.

    2002-03-01

    modelling of the interfacial area construction, can be directly applied to the design of accelerator driven systems (ADS). However, the transfer from the two-phase system water / steam to the system lead-bismuth / nitrogen still has to be done. Thus, the results of the POOLTHY Project are partly of direct relevance to the Programme on Partitioning and Transmutation and especially to the PDS-XADS Project. (orig.)

  15. Steam generator life management

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

    1998-01-01

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermal hydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. This report shows how recent advances in cleaning technology are integrated into a life management strategy, discusses downcomer flow measurement as a means of monitoring steam generator condition, and describes recent advances in hideout return as a life management tool. The research and development program, as well as operating experience, has identified

  16. Two-phase flow measurement based on oblique laser scattering

    Science.gov (United States)

    Vendruscolo, Tiago P.; Fischer, Robert; Martelli, Cícero; Rodrigues, Rômulo L. P.; Morales, Rigoberto E. M.; da Silva, Marco J.

    2015-07-01

    Multiphase flow measurements play a crucial role in monitoring productions processes in many industries. To guarantee the safety of processes involving multiphase flows, it is important to detect changes in the flow conditions before they can cause damage, often in fractions of seconds. Here we demonstrate how the scattering pattern of a laser beam passing a two-phase flow under an oblique angle to the flow direction can be used to detect derivations from the desired flow conditions in microseconds. Applying machine-learning techniques to signals obtained from three photo-detectors we achieve a compact, versatile, low-cost sensor design for safety applications.

  17. Modeling of annular two-phase flow using a unified CFD approach

    Energy Technology Data Exchange (ETDEWEB)

    Li, Haipeng, E-mail: haipengl@kth.se; Anglart, Henryk, E-mail: henryk@kth.se

    2016-07-15

    Highlights: • Annular two-phase flow has been modeled using a unified CFD approach. • Liquid film was modeled based on a two-dimensional thin film assumption. • Both Eulerian and Lagrangian methods were employed for the gas core flow modeling. - Abstract: A mechanistic model of annular flow with evaporating liquid film has been developed using computational fluid dynamics (CFD). The model is employing a separate solver with two-dimensional conservation equations to predict propagation of a thin boiling liquid film on solid walls. The liquid film model is coupled to a solver of three-dimensional conservation equations describing the gas core, which is assumed to contain a saturated mixture of vapor and liquid droplets. Both the Eulerian–Eulerian and the Eulerian–Lagrangian approach are used to describe the droplet and vapor motion in the gas core. All the major interaction phenomena between the liquid film and the gas core flow have been accounted for, including the liquid film evaporation as well as the droplet deposition and entrainment. The resultant unified framework for annular flow has been applied to the steam-water flow with conditions typical for a Boiling Water Reactor (BWR). The simulation results for the liquid film flow rate show good agreement with the experimental data, with the potential to predict the dryout occurrence based on criteria of critical film thickness or critical film flow rate.

  18. Modeling of annular two-phase flow using a unified CFD approach

    International Nuclear Information System (INIS)

    Li, Haipeng; Anglart, Henryk

    2016-01-01

    Highlights: • Annular two-phase flow has been modeled using a unified CFD approach. • Liquid film was modeled based on a two-dimensional thin film assumption. • Both Eulerian and Lagrangian methods were employed for the gas core flow modeling. - Abstract: A mechanistic model of annular flow with evaporating liquid film has been developed using computational fluid dynamics (CFD). The model is employing a separate solver with two-dimensional conservation equations to predict propagation of a thin boiling liquid film on solid walls. The liquid film model is coupled to a solver of three-dimensional conservation equations describing the gas core, which is assumed to contain a saturated mixture of vapor and liquid droplets. Both the Eulerian–Eulerian and the Eulerian–Lagrangian approach are used to describe the droplet and vapor motion in the gas core. All the major interaction phenomena between the liquid film and the gas core flow have been accounted for, including the liquid film evaporation as well as the droplet deposition and entrainment. The resultant unified framework for annular flow has been applied to the steam-water flow with conditions typical for a Boiling Water Reactor (BWR). The simulation results for the liquid film flow rate show good agreement with the experimental data, with the potential to predict the dryout occurrence based on criteria of critical film thickness or critical film flow rate.

  19. Two phase flow problems in power station boilers

    International Nuclear Information System (INIS)

    Firman, E.C.

    1974-01-01

    The paper outlines some of the waterside thermal and hydrodynamic phenomena relating to design and operation of large boilers in central power stations. The associated programme of work is described with an outline of some results already obtained. By way of introduction, the principal features of conventional and nuclear drum boilers and once-through nuclear heat exchangers are described in so far as they pertain to this area of work. This is followed by discussion of the relevant physical phenomena and problems which arise. For example, the problem of steam entrainment from the drum into the tubes connecting it to the furnace wall tubes is related to its effects on circulation and possible mechanisms of tube failure. Other problems concern the transient associated with start-up or low load operation of plant. The requirement for improved mathematical representation of steady and dynamic performance is mentioned together with the corresponding need for data on heat transfer, pressure loss, hydrodynamic stability, consequences of deposits, etc. The paper concludes with reference to the work being carried out within the C.E.G.B. in relation to the above problems. The facilities employed and the specific studies being made on them are described: these range from field trials on operational boilers to small scale laboratory investigations of underlying two phase flow mechanisms and include high pressure water rigs and a freon rig for simulation studies

  20. Two-phase ozonation of chlorinated organics

    International Nuclear Information System (INIS)

    Bhattacharyya, D.; Freshour, A.; West, D.

    1995-01-01

    In the last few years the amount of research being conducted in the field of single-phase ozonation has grown extensively. However, traditional aqueous-phase ozonation systems are limited by a lack of selective oxidation potential, low ozone solubility in water, and slow intermediate decomposition rates. Furthermore, ozone may decompose before it can be utilized for pollutant destruction since ozone can be highly unstable in aqueous solutions. Naturally occurring compounds such as NaHCO 3 also affect ozone reactions by inhibiting the formation of OH-free radicals. To compensate for these factors, excess ozone is typically supplied to a reactor. Since ozone generation requires considerable electric power consumption (16 - 24 kWh/kg of O 3 ), attempts to enhance the ozone utilization rate and stability should lead to more efficient application of this process to hazardous waste treatment. To improve the process, ozonation may be more efficiently carried out in a two-phase system consisting of an inert solvent (saturated with O 3 ) contacted with an aqueous phase containing pollutants. The non-aqueous phase must meet the following criteria: (1) non-toxic, (2) very low vapor pressure, (3) high density (for ease of separation), (4) complete insolubility in water, (5) reusability, (6) selective pollutant extractability, (7) high oxidant solubility, and (8) extended O 3 stability. Previously published studies (1) have indicated that a number of fluorinated hydrocarbon compounds fit these criteria. For this project, FC40 (a product of 3M Co.) was chosen due to its low vapor pressure (3 mm Hg) and high specific gravity (1.9). The primary advantages of the FC40 solvent are that it is non-toxic, reusable, has an ozone solubility 10 times that of water, and that 85 % of the ozone remains in the solvent even after 2 hours. This novel two-phase process has been utilized to study the rapid destruction of organic chlorine compounds and organic mixtures

  1. Characteristics of low-mass-velocity vertical gas-liquid two-phase flow

    International Nuclear Information System (INIS)

    Adachi, Hiromichi; Abe, Yutaka; Kimura, Ko-ji

    1995-01-01

    In the present paper, characteristics of low mass velocity two-phase flow was analyzed based on a concept that pressure energy of two-phase flow is converted into acceleration work, gravitational work and frictional work, and the pressure energy consumption rate should be minimum at the stable two-phase flow condition. Experimental data for vertical upward air-water two-phase flow at atmospheric pressure was used to verify this concept and the turbulent model used in this method is optimized with the data. (author)

  2. Fluid dynamics of cryogenic two-phase flows

    International Nuclear Information System (INIS)

    Verfondern, K.; Jahn, W.

    2004-01-01

    The objective of this study was to examine the flow behavior of a methane hydrate/methane-liquid hydrogen dispersed two-phase fluid through a given design of a moderator chamber for the ESS target system. The calculations under simplified conditions, e.g., taking no account of heat input from outside, have shown that the computer code used, CFX, was able to simulate the behavior of the two-phase flow through the moderator chamber, producing reasonable results up to a certain level of the solid phase fraction, that allowed a continuous flow process through the chamber. Inlet flows with larger solid phase fractions than 40 vol% were found to be a ''problem'' for the computer code. From the computer runs based on fractions between 20 and 40 vol%, it was observed that with increasing solid phase fraction at the inlet, the resulting flow pattern revealed a strong tendency for blockage within the chamber, supported by the ''heavy weight'' of the pellets compared to the carrying liquid. Locations which are prone to the development of such uneven flow behavior are the areas around the turning points in the semispheres and near the exit of the moderator. The considered moderator chamber with horizontal inlet and outlet flow for a solid-liquid two-phase fluid does not seem to be an appropriate design. (orig.)

  3. Numerical simulation of two phase flows in heat exchangers; Simulation numerique des ecoulements diphasiques dans les echangeurs

    Energy Technology Data Exchange (ETDEWEB)

    Grandotto Biettoli, M

    2006-04-15

    The report presents globally the works done by the author in the thermohydraulic applied to nuclear reactors flows. It presents the studies done to the numerical simulation of the two phase flows in the steam generators and a finite element method to compute these flows. (author)

  4. Interfacial structures in downward two-phase bubbly flow

    International Nuclear Information System (INIS)

    Paranjape, S.S.; Kim, S.; Ishii, M.; Kelly, J.

    2003-01-01

    Downward two-phase flow was studied considering its significance in view of Light Water Reactor Accidents (LWR) such as Loss of Heat Sink (LOHS) by feed water loss or secondary pipe break. The flow studied, was an adiabatic, air-water, co-current, vertically downward two-phase flow. The experimental test sections had internal hydraulic diameters of 25.4 mm and 50.8 mm. Flow regime map was obtained using the characteristic signals obtained from an impedance void meter, employing neural network based identification methodology to minimize the subjective judgment in determining the flow regimes. A four sensor conductivity probe was used to measure the local two phase flow parameters, which characterize the interfacial structures. The local time averaged two-phase flow parameters measured were: void fraction (α), interfacial area concentration (a i ), bubble velocity (v g ), and Sauter mean diameter (D Sm ). The flow conditions were from the bubbly flow regime. The local profiles of these parameters as well as their axial development revealed the nature of the interfacial structures and the bubble interaction mechanisms occurring in the flow. Furthermore, this study provided a good database for the development of the interfacial area transport equation, which dynamically models the changes in the interfacial area along the flow field. An interfacial area transport equation was developed for downward flow based on that developed for the upward flow, with certain modifications in the bubble interaction terms. The area averaged values of the interfacial area concentration were compared with those predicted by the interfacial area transport model. (author)

  5. Design and development of drag-disc flowmeter for measurement of transient two-phase flow

    International Nuclear Information System (INIS)

    Sreenivas Rao, G.; Kukreja, V.; Dolas, P.K.; Venkat Raj, V.

    1990-01-01

    Experiments have been carried out to test the suitability of drag-disc flowmeter for measuring two-phase flow. Calibration tests carried out under single-phase and two-phase flow conditions have confirmed the suitability of the drag-disc flowmeter. The experimental work and the results obtained are presented and discussed in the paper. (author). 3 refs., 6 figs

  6. Characterization of horizontal air–water two-phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Kong, Ran; Kim, Seungjin, E-mail: skim@psu.edu

    2017-02-15

    Highlights: • A visualization study is performed to develop flow regime map in horizontal flow. • Database in horizontal bubbly flow is extended using a local conductivity probe. • Frictional pressure drop analysis is performed in horizontal bubbly flow. • Drift flux analysis is performed in horizontal bubbly flow. - Abstract: This paper presents experimental studies performed to characterize horizontal air–water two-phase flow in a round pipe with an inner diameter of 3.81 cm. A detailed flow visualization study is performed using a high-speed video camera in a wide range of two-phase flow conditions to verify previous flow regime maps. Two-phase flows are classified into bubbly, plug, slug, stratified, stratified-wavy, and annular flow regimes. While the transition boundaries identified in the present study compare well with the existing ones (Mandhane et al., 1974) in general, some discrepancies are observed for bubbly-to-plug/slug, and plug-to-slug transition boundaries. Based on the new transition boundaries, three additional test conditions are determined in horizontal bubbly flow to extend the database by Talley et al. (2015a). Various local two-phase flow parameters including void fraction, interfacial area concentration, bubble velocity, and bubble Sauter mean diameter are obtained. The effects of increasing gas flow rate on void fraction, bubble Sauter mean diameter, and bubble velocity are discussed. Bubbles begin to coalesce near the gas–liquid layer instead of in the highly packed region when gas flow rate increases. Using all the current experimental data, two-phase frictional pressure loss analysis is performed using the Lockhart–Martinelli method. It is found that the coefficient C = 24 yields the best agreement with the data with the minimum average difference. Moreover, drift flux analysis is performed to predict void-weighted area-averaged bubble velocity and area-averaged void fraction. Based on the current database, functional

  7. An investigation of nucleating flows of steam in a cascade of turbine blading: Effect of overall pressure ratios

    International Nuclear Information System (INIS)

    Bakhtar, F.; Savage, R.A.

    1993-01-01

    In the course of expansion of steam in turbines the state path crosses the saturation line and the fluid becomes a two-phase mixture. To reproduce turbine nucleating and wet conditions realistically requires a supply of supercooled steam which can be obtained under blow down conditions. An experimental short duration cascade tunnel working on this principle has been constructed. The blade profile studied is that of a typical nozzle The paper is one of a set and describes the surface pressure measurements carried out to investigate the effect of the overall pressure ratio on the performance of the blade

  8. Structure of the gas-liquid annular two-phase flow in a nozzle section

    International Nuclear Information System (INIS)

    Yoshida, Kenji; Kataoka, Isao; Ohmori, Syuichi; Mori, Michitsugu

    2006-01-01

    Experimental studies on the flow behavior of gas-liquid annular two-phase flow passing through a nozzle section were carried out. This study is concerned with the central steam jet injector for a next generation nuclear reactor. In the central steam jet injector, steam/water annular two-phase flow is formed at the mixing nozzle. To make an appropriate design and to establish the high-performance steam injector system, it is very important to accumulate the fundamental data of the thermo-hydro dynamic characteristics of annular flow passing through a nozzle section. On the other hand, the transient behavior of multiphase flow, in which the interactions between two-phases occur, is one of the most interesting scientific issues and has attracted research attention. In this study, the transient gas-phase turbulence modification in annular flow due to the gas-liquid phase interaction is experimentally investigated. The annular flow passing through a throat section is under the transient state due to the changing cross sectional area of the channel and resultantly the superficial velocities of both phases are changed compared with a fully developed flow in a straight pipe. The measurements for the gas-phase turbulence were precisely performed by using a constant temperature hot-wire anemometer, and made clear the turbulence structure such as velocity profiles, fluctuation velocity profiles. The behavior of the interfacial waves in the liquid film flow such as the ripple or disturbance waves was also observed. The measurements for the liquid film thickness by the electrode needle method were also performed to measure the base film thickness, mean film thickness, maximum film thickness and wave height of the ripple or the disturbance waves. (author)

  9. Steam drums

    International Nuclear Information System (INIS)

    Crowder, R.

    1978-01-01

    Steam drums are described that are suitable for use in steam generating heavy water reactor power stations. They receive a steam/water mixture via riser headers from the reactor core and provide by means of separators and driers steam with typically 0.5% moisture content for driving turbines. The drums are constructed as prestressed concrete pressure vessels in which the failure of one or a few of the prestressing elements does not significantly affect the overall strength of the structure. The concrete also acts as a radiation shield. (U.K.)

  10. The Concept of Steam Pressure Control by Changing the Feedwater Flow during Heatup Operation for an Integral Reactor with a Once-Through Steam Generator

    International Nuclear Information System (INIS)

    Seo, Jae Kwang; Choi, Ki Yong; Kang, Han Ok; Kim, Young In; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    The design features of a once-through steam generator (OTSG) for an integral reactor are significantly different from the commercial U-tube type steam generator from several aspects such as the general arrangement, size, operation conditions, and so on. Therefore a sufficient understanding of the thermal-hydraulic characteristics of the OTSG is essential for the design of the nuclear steam supply system (NSSS) and the power conversion system (PCS). It is also necessary to develop operation procedures complying to the unique design features of the OTSG of interest. The OTSG is sized to produce a sufficiently superheated steam during a normal power operation and therefore the secondary system can be simple relative to that of the other types of steam generators. For the plant adopting the OTSG, the steam pressure in the secondary circuit (tube side of the OTSG) is controlled to be constant during a normal power operation. Constant steam pressure is realized by regulating the control valve on the main steam line dedicated for this purpose. However during a heatup operation, at which the fluid state at the exit of the OTSG is a single phase hot water or two phases, it is not proper to use the control valve on the main steam line due to a control problem at low and multi-phase flow conditions and possibly an erosion problem. For these reasons, another dedicated line called a startup cooling line is used during a heatup condition. There may be several operational conditions for the secondary fluid required to pass through during heatup operation, depending on the design of the PCS. In general, there are two conditions: One is a condition for a vacuum operation for the condenser and another is an entry condition for a steam pressure control operation for an auxiliary power system. In this study, the concept of using a simple startup cooling line with a fixed flow resistance and changing the feedwater flow for the pressure control of the PCS during a heatup period are

  11. Developing two-phase flow modelling concepts for rock fractures

    Energy Technology Data Exchange (ETDEWEB)

    Keto, V. (Fortum Nuclear Services Oy, Espoo (Finland))

    2010-01-15

    The Finnish nuclear waste disposal company, Posiva Oy, is planning an underground repository for spent nuclear fuel to be constructed on the island of Olkiluoto on the south-west coast of Finland. One element of the site investigations conducted at Olkiluoto is the excavation of the underground rock characterisation facility (ONKALO) that will be extended to the final disposal depth (approximately -400 m). The bedrock around the excavated tunnel volume is fully saturated with groundwater, which water commonly contains a mixture of dissolved gases. These gases remain dissolved due to the high hydrostatic pressure. During tunnel excavation work the natural hydrostatic pressure field is disturbed and the water pressure will decrease close to the atmospheric pressure in the immediate vicinity of the tunnel. During this pressure drop two-phase flow conditions (combined flow of both water and gas) may develop in the vicinity of the underground opening, as the dissolved gas is exsoluted under the low pressure (the term exsolution refers here to release of the dissolved gas molecules from the water phase into a separate gas phase). This report steers towards concept development for numerical two-phase flow modeling for fractured rock. The focus is on the description of gas phase formation process under disturbed hydraulic conditions by exsolution of dissolved gases from groundwater, and on understanding the effects of a possibly formed gas phase on groundwater flow conditions in rock fractures. A mathematical model of three mutually coupled nonlinear partial differential equations for two-phase flow is presented and corresponding constitutional relationships are introduced and discussed. Illustrative numerical simulations are performed in a simplified setting using COMSOL Multiphysics 3.5a - software package. Shortcomings and conceptual problems are discussed. (orig.)

  12. Developing two-phase flow modelling concepts for rock fractures

    International Nuclear Information System (INIS)

    Keto, V.

    2010-01-01

    The Finnish nuclear waste disposal company, Posiva Oy, is planning an underground repository for spent nuclear fuel to be constructed on the island of Olkiluoto on the south-west coast of Finland. One element of the site investigations conducted at Olkiluoto is the excavation of the underground rock characterisation facility (ONKALO) that will be extended to the final disposal depth (approximately -400 m). The bedrock around the excavated tunnel volume is fully saturated with groundwater, which water commonly contains a mixture of dissolved gases. These gases remain dissolved due to the high hydrostatic pressure. During tunnel excavation work the natural hydrostatic pressure field is disturbed and the water pressure will decrease close to the atmospheric pressure in the immediate vicinity of the tunnel. During this pressure drop two-phase flow conditions (combined flow of both water and gas) may develop in the vicinity of the underground opening, as the dissolved gas is exsoluted under the low pressure (the term exsolution refers here to release of the dissolved gas molecules from the water phase into a separate gas phase). This report steers towards concept development for numerical two-phase flow modeling for fractured rock. The focus is on the description of gas phase formation process under disturbed hydraulic conditions by exsolution of dissolved gases from groundwater, and on understanding the effects of a possibly formed gas phase on groundwater flow conditions in rock fractures. A mathematical model of three mutually coupled nonlinear partial differential equations for two-phase flow is presented and corresponding constitutional relationships are introduced and discussed. Illustrative numerical simulations are performed in a simplified setting using COMSOL Multiphysics 3.5a - software package. Shortcomings and conceptual problems are discussed. (orig.)

  13. Two-Phase Quality/Flow Meter

    Science.gov (United States)

    Moerk, J. Steven (Inventor); Youngquist, Robert C. (Inventor); Werlink, Rudy J. (Inventor)

    1999-01-01

    A quality and/or flow meter employs a capacitance probe assembly for measuring the dielectric constant of flow stream, particularly a two-phase flow stream including liquid and gas components.ne dielectric constant of the flow stream varies depending upon the volume ratios of its liquid and gas components, and capacitance measurements can therefore be employed to calculate the quality of the flow, which is defined as the volume ratio of liquid in the flow to the total volume ratio of gas and liquid in the flow. By using two spaced capacitance sensors, and cross-correlating the time varying capacitance values of each, the velocity of the flow stream can also be determined. A microcontroller-based processing circuit is employed to measure the capacitance of the probe sensors.The circuit employs high speed timer and counter circuits to provide a high resolution measurement of the time interval required to charge each capacitor in the probe assembly. In this manner, a high resolution, noise resistant, digital representation of each of capacitance value is obtained without the need for a high resolution A/D converter, or a high frequency oscillator circuit. One embodiment of the probe assembly employs a capacitor with two ground plates which provide symmetry to insure that accurate measurements are made thereby.

  14. Two-phase flux simulations by robots

    International Nuclear Information System (INIS)

    Barrera, F.D.

    1997-01-01

    Two-Phase flow systems are studied following the statistical formulation, which takes into account the bubble population balances. This is done by means of automata simulation. Geometrical automata are associated to the dispersed phase, and are represented by discs on the plane, resembling bubbles moving in a fluid environment. Following pre-determined rules, the automata evolve, and useful statistical information about their interaction is obtained. This information is applied in the present work to study the mechanisms that induce bubble coalescence. Models for one and two sized automata are presented. It was found that in the case of the model for one size, the probability of interaction among bubbles and the pair correlation function depends not only on the void fraction, but also on the number of elements of the dispersed phase. A correlation for the collision probability between two bubbles is obtained, and this result was extended to the pair correlation function. For the case of systems with two characteristic sizes, a model was formulated for analyzing the interaction among bubbles of the two groups. The interaction of bubbles for one and two sized systems were related by a symmetry factor, which shows the dependence of the interaction among bubbles with the size distribution. By means of the automata simulation, the phenomena of bubble confinement and screening were characterized. It was found that the first phenomenon is stronger in systems with greater distance among bubbles, and that the second effect increases with void fraction and bubble number. (author)

  15. Numerical calculation of two-phase flows

    International Nuclear Information System (INIS)

    Travis, J.R.; Harlow, F.H.; Amsden, A.A.

    1975-06-01

    The theoretical study of time-varying two-phase flow problems in several space dimensions introduces such a complicated set of coupled nonlinear partial differential equations that numerical solution procedures for high-speed computers are required in almost all but the simplest examples. Efficient attainment of realistic solutions for practical problems requires a finite- difference formulation that is simultaneously implicit in the treatment of mass convection, equations of state, and the momentum coupling between phases. Such a method is described, the equations on which it is based are discussed, and its properties are illustrated by means of examples. In particular, the capability for calculating physical instabilities and other time-varying dynamics, at the same time avoiding numerical instability is emphasized. The computer code is applicable to problems in reactor safety analysis, the dynamics of fluidized dust beds, raindrops or aerosol transport, and a variety of similar circumstances, including the effects of phase transitions and the release of latent heat or chemical energy. (U.S.)

  16. Condensation in a two-phase pool

    International Nuclear Information System (INIS)

    Duffey, R.B.; Hughes, E.D.

    1991-01-01

    We consider the case of vapor condensation in a liquid pool, when the heat transfer is controlled by heat losses through the walls. The analysis is based on drift flux theory for phase separation in the pool, and determines the two-phase mixture height for the pool. To our knowledge this is the first analytical treatment of this classic problem that gives an explicit result, previous work having established the result for the evaporative case. From conservation of mass and energy in a one-dimensional steady flow, together with a void relation between the liquid and vapor fluxes, we determine the increase in the mixture level from the base level of the pool. It can be seen that the thermal and hydrodynamic influences are separable. Thus, the thermal influence of the wall heat transfer appears through its effect on the condensing length L*, so that at high condensation rates the pool is all liquid, and at low rates overflows (the level swell or foaming effect). Similarly, the phase separation effect hydrodynamically determines the height via the relative velocity of the mixture to the entering flux. We examine some practical applications of this result to level swell in condensing flows, and also examine some limits in ideal cases

  17. Post Analysis of Two Phase Natural Circulation Mass Flow Rate for CE-PECS

    Energy Technology Data Exchange (ETDEWEB)

    Park, R. J.; Ha, K. S.; Rhee, B. W.; Kim, H. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The coolant in the inclined channel absorbs the decay heat and sensible heat transferred from the corium through the structure of the core catcher body and flows up to the pool as a two phase mixture. On the other hand, some of the pool water will flow into the inlet of the downcomer piping, and will flow into the inclined cooling channel of the core catcher by gravity. The engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting. To maintain the integrity of the ex-vessel core catcher, however, it is required that the coolant be circulated at a rate along the inclined cooling channel sufficient to avoid CHF (Critical Heat Flux) on the heating surface of the cooling channel. In this study, post simulations of two phase natural circulation in the CEPECS have been performed to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. Post simulations of two phase natural circulation in the CE-PECS have been conducted to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is approximately 8.7 kg/s in the base case.

  18. Post Analysis of Two Phase Natural Circulation Mass Flow Rate for CE-PECS

    International Nuclear Information System (INIS)

    Park, R. J.; Ha, K. S.; Rhee, B. W.; Kim, H. Y.

    2015-01-01

    The coolant in the inclined channel absorbs the decay heat and sensible heat transferred from the corium through the structure of the core catcher body and flows up to the pool as a two phase mixture. On the other hand, some of the pool water will flow into the inlet of the downcomer piping, and will flow into the inclined cooling channel of the core catcher by gravity. The engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting. To maintain the integrity of the ex-vessel core catcher, however, it is required that the coolant be circulated at a rate along the inclined cooling channel sufficient to avoid CHF (Critical Heat Flux) on the heating surface of the cooling channel. In this study, post simulations of two phase natural circulation in the CEPECS have been performed to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. Post simulations of two phase natural circulation in the CE-PECS have been conducted to evaluate two phase flow characteristics and the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is approximately 8.7 kg/s in the base case

  19. Flooding and flow reversal of two-phase annular flow

    International Nuclear Information System (INIS)

    Asahi, Y.

    1978-01-01

    The flooding and flow reversal conditions of two-phase annular flow are mathematically defined in terms of a characteristic function representing a force balance. Sufficiently below the flooding point in counter-current flow, the interface is smooth and the characteristic equation reduces to the Nusselt relationship. Just below flooding point and above the flow reversal point in cocurrent flow, the interface is 'wavy', so that the interfacial shear effect plays an important role. The theoretical analysis is compared with experimental results by others. It is suggested that the various length effects which have been experimentally observed may be accounted for by the spatial variation of the droplet entrainment. (Auth.)

  20. Current capabilities of transient two-phase flow instruments

    International Nuclear Information System (INIS)

    Solbrig, C.W.; Kondic, N.N.

    1979-01-01

    The measurement of two phase flow phenomena in transient conditions representative of a Loss-of-Coolant Accident requires the use of sophisticated instruments and the further development of other instruments. Measurements made in large size pipes are often flow regime dependent. The flow regimes encountered depend upon the system geometry, transient effects, heat transfer, etc. The geometries in which these measurements must be made, the instruments which are currently used, new instruments being developed, the facilities used to calibrate these instruments, and the improvements which must be made to measurement capabilities are described

  1. Dynamic modelling for two-phase flow systems

    International Nuclear Information System (INIS)

    Guerra, M.A.

    1991-06-01

    Several models for two-phase flow have been studied, developing a thermal-hydraulic analysis code with one of these models. The program calculates, for one-dimensional cases with variable flow area, the transient behaviour of system process variables, when the boundary conditions (heat flux, flow rate, enthalpy and pressure) are functions of time. The modular structure of the code, eases the program growth. In fact, the present work is the basis for a general purpose accident and transient analysis code in nuclear reactors. Code verification has been made against RETRAN-02 results. Satisfactory results have been achieved with the present version of the code. (Author) [es

  2. Interfacial area measurements in two-phase flow

    International Nuclear Information System (INIS)

    Veteau, J.-M.

    1979-08-01

    A thorough understanding of two-phase flow requires the accurate measurement of the time-averaged interfacial area per unit volume (also called the time-averaged integral specific area). The so-called 'specific area' can be estimated by several techniques described in the literature. These different methods are reviewed and the flow conditions which lead to a rigourous determination of the time-averaged integral specific area are clearly established. The probe technique, involving local measurements seems very attractive because of its large range of application [fr

  3. Numerical simulation of two phase flows in heat exchangers

    International Nuclear Information System (INIS)

    Grandotto Biettoli, M.

    2006-04-01

    The author gives an overview of his research activity since 1981. He first gives a detailed presentation of properties and equations of two-phase flows in heat exchangers, and of their mathematical and numerical investigation: semi-local equations (mass conservation, momentum conservation and energy conservation), homogenized conservation equations (mass, momentum and enthalpy conservation, boundary conditions), equation closures, discretization, resolution algorithm, computational aspects and applications. Then, he reports the works performed in the field of turbulent flows, hyperbolic methods, low Mach methods, the Neptune project, and parallel computing

  4. Steam 80 steam generator instrumentation

    International Nuclear Information System (INIS)

    Carson, W.H.; Harris, H.H.

    1980-01-01

    This paper describes two special instrumentation packages in an integral economizer (preheater) steam generator of one of the first System 80 plants scheduled to go into commercial operation. The purpose of the instrumentation is to obtain accurate operating information from regions of the secondary side of the steam generator inaccessible to normal plant instrumentation. In addition to verification of the System 80 steam generator design predictions, the data obtained will assist in verification of steam generator thermal/hydraulic computer codes developed for generic use in the industry

  5. Experimental study of two-phase natural circulation circuit

    Energy Technology Data Exchange (ETDEWEB)

    Lemos, Wanderley Freitas; Su, Jian, E-mail: wlemos@lasme.coppe.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Faccini, Jose Luiz Horacio, E-mail: faccini@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), RIo de Janeiro, RJ (Brazil). Lab. de Termo-Hidraulica Experimental

    2012-07-01

    This paper reports an experimental study on the behavior of fluid flow in natural circulation under single-and two-phase flow conditions. The natural circulation circuit was designed based on concepts of similarity and scale in proportion to the actual operating conditions of a nuclear reactor. This test equipment has similar performance to the passive system for removal of residual heat presents in Advanced Pressurized Water Reactors (A PWR). The experiment was carried out by supplying water to primary and secondary circuits, as well as electrical power resistors installed inside the heater. Power controller has available to adjust the values for supply of electrical power resistors, in order to simulate conditions of decay of power from the nuclear reactor in steady state. Data acquisition system allows the measurement and control of the temperature at different points by means of thermocouples installed at several points along the circuit. The behavior of the phenomenon of natural circulation was monitored by a software with graphical interface, showing the evolution of temperature measurement points and the results stored in digital format spreadsheets. Besides, the natural circulation flow rate was measured by a flowmeter installed on the hot leg. A flow visualization technique was used the for identifying vertical flow regimes of two-phase natural circulation. Finally, the Reynolds Number was calculated for the establishment of a friction factor correlation dependent on the scale geometrical length, height and diameter of the pipe. (author)

  6. Experimental study of two-phase natural circulation circuit

    International Nuclear Information System (INIS)

    Lemos, Wanderley Freitas; Su, Jian; Faccini, Jose Luiz Horacio

    2012-01-01

    This paper reports an experimental study on the behavior of fluid flow in natural circulation under single-and two-phase flow conditions. The natural circulation circuit was designed based on concepts of similarity and scale in proportion to the actual operating conditions of a nuclear reactor. This test equipment has similar performance to the passive system for removal of residual heat presents in Advanced Pressurized Water Reactors (A PWR). The experiment was carried out by supplying water to primary and secondary circuits, as well as electrical power resistors installed inside the heater. Power controller has available to adjust the values for supply of electrical power resistors, in order to simulate conditions of decay of power from the nuclear reactor in steady state. Data acquisition system allows the measurement and control of the temperature at different points by means of thermocouples installed at several points along the circuit. The behavior of the phenomenon of natural circulation was monitored by a software with graphical interface, showing the evolution of temperature measurement points and the results stored in digital format spreadsheets. Besides, the natural circulation flow rate was measured by a flowmeter installed on the hot leg. A flow visualization technique was used the for identifying vertical flow regimes of two-phase natural circulation. Finally, the Reynolds Number was calculated for the establishment of a friction factor correlation dependent on the scale geometrical length, height and diameter of the pipe. (author)

  7. Investigation of grid-enhanced two-phase convective heat transfer in the dispersed flow film boiling regime

    International Nuclear Information System (INIS)

    Miller, D.J.; Cheung, F.B.; Bajorek, S.M.

    2013-01-01

    Highlights: • Experiments were done in the RBHT facility to study the droplet flow in rod bundle. • The presence of a droplet field was found to greatly enhance heat transfer. • A second-stage augmentation was observed downstream of a spacer grid. • This augmentation is due to the breakup of liquid ligaments downstream of the grid. - Abstract: A two-phase dispersed droplet flow investigation of the grid-enhanced heat transfer augmentation has been done using steam cooling with droplet injection experimental data obtained from the Penn State/NRC Rod Bundle Heat Transfer (RBHT) facility. The RBHT facility is a vertical, full length, 7 × 7-rod bundle heat transfer facility having 45 electrically heated fuel rod simulators of 9.5 mm (0.374-in.) diameter on a 12.6 mm (0.496-in.) pitch which simulates a portion of a PWR fuel assembly. The facility operates at low pressure, up to 4 bars (60 psia) and has over 500 channels of instrumentation including heater rod thermocouples, spacer grid thermocouples, closely-spaced differential pressure cells along the test section, several fluid temperature measurements within the rod bundle flow area, inlet and exit flows, absolute pressure, and the bundle power. A series of carefully controlled and well instrumented steam cooling with droplet injection experiments were performed over a range of Reynolds numbers and droplet injection flow rates. The experimental results were analyzed to obtain the axial variation of the local heat transfer coefficients along the rod bundle. At the spacer grid location, the flow was found to be substantially disrupted, with the hydrodynamic and thermal boundary layers undergoing redevelopment. Owing to this flow restructuring, the heat transfer downstream of a grid spacer was found to be augmented above the fully developed flow heat transfer as a result of flow disruption induced by the grid. Furthermore, the presence of a droplet field further enhanced the heat transfer as compared to single

  8. Selling steam

    International Nuclear Information System (INIS)

    Zimmer, M.J.; Goodwin, L.M.

    1991-01-01

    This article addresses the importance of steam sales contract is in financing cogeneration facilities. The topics of the article include the Public Utility Regulatory Policies Act provisions and how they affect the marketing of steam from qualifying facilities, the independent power producers market shift, and qualifying facility's benefits

  9. Steam generator

    International Nuclear Information System (INIS)

    Fenet, J.-C.

    1980-01-01

    Steam generator particularly intended for use in the coolant system of a pressurized water reactor for vaporizing a secondary liquid, generally water, by the primary cooling liquid of the reactor and comprising special arrangements for drying the steam before it leaves the generator [fr

  10. Steam condenser

    International Nuclear Information System (INIS)

    Masuda, Fujio

    1980-01-01

    Purpose: To enable safe steam condensation by providing steam condensation blades at the end of a pipe. Constitution: When high temperature high pressure steam flows into a vent pipe having an opening under water in a pool or an exhaust pipe or the like for a main steam eacape safety valve, non-condensable gas filled beforehand in the steam exhaust pipe is compressed, and discharged into the water in the pool. The non-condensable gas thus discharged from the steam exhaust pipe is introduced into the interior of the hollow steam condensing blades, is then suitably expanded, and thereafter exhausted from a number of exhaust holes into the water in the pool. In this manner, the non-condensable gas thus discharged is not directly introduced into the water in the pool, but is suitable expanded in the space of the steam condensing blades to suppress extreme over-compression and over-expansion of the gas so as to prevent unstable pressure vibration. (Yoshihara, H.)

  11. Interaction between local parameters of two-phase flow and random forces on a cylinder

    International Nuclear Information System (INIS)

    Sylviane Pascal-Ribot; Yves Blanchet; Franck Baj; Phillippe Piteau

    2005-01-01

    Full text of publication follows: In the frame of assessments of steam generator tube bundle vibrations, a study was conducted in order to investigate the effects of an air/water flow on turbulent buffeting forces induced on a cylinder. The main purpose is to relate the physical parameters characterizing an air/water two-phase crossflow with the structural loading of a fixed cylindrical tube. In this first approach, the experiments are carried out in a rectangular acrylic test section supplied with a vertical upward bubbly flow. This flow is transversally impeded by a fixed rigid 12,15 mm diameter cylinder. Different turbulence grids are used in order to modify two-phase characteristics such as bubble diameter, void fraction profile, fluctuation parameters. Preliminarily, a dimensional analysis of fluid-structure interaction under two-phase turbulent solicitations has enabled to identify a list of physically relevant variables which must be measured to evaluate the random forces. The meaning of these relevant parameters as well as the effect of flow patterns are discussed. Direct measurements of two-phase flow parameters are performed simultaneously with measurements of forces exerted on the cylinder. The main descriptive parameters of a two-phase flow are measured using a bi-optical probe, in particular void fraction profiles, interfacial velocities, bubble diameters, void fraction fluctuations. In the same time, the magnitude of random forces caused by two-phase flow is measured with a force transducer. A thorough analysis of the experimental data is then undertaken in order to correlate physical two-phase mechanisms with the random forces exerted on the cylinder. The hypotheses made while applying the dimensional analysis are verified and their pertinence is discussed. Finally, physical parameters involved in random buffeting forces applied on a transverse tube are proposed to scale the spectral magnitude of these forces and comparisons with other authors

  12. Two-phase Heating in Flaring Loops

    Science.gov (United States)

    Zhu, Chunming; Qiu, Jiong; Longcope, Dana W.

    2018-03-01

    We analyze and model a C5.7 two-ribbon solar flare observed by the Solar Dynamics Observatory, Hinode, and GOES on 2011 December 26. The flare is made of many loops formed and heated successively over one and half hours, and their footpoints are brightened in the UV 1600 Å before enhanced soft X-ray and EUV missions are observed in flare loops. Assuming that anchored at each brightened UV pixel is a half flaring loop, we identify more than 6700 half flaring loops, and infer the heating rate of each loop from the UV light curve at the footpoint. In each half loop, the heating rate consists of two phases: intense impulsive heating followed by a low-rate heating that is persistent for more than 20 minutes. Using these heating rates, we simulate the evolution of their coronal temperatures and densities with the model of the “enthalpy-based thermal evolution of loops.” In the model, suppression of thermal conduction is also considered. This model successfully reproduces total soft X-ray and EUV light curves observed in 15 passbands by four instruments GOES, AIA, XRT, and EVE. In this flare, a total energy of 4.9 × 1030 erg is required to heat the corona, around 40% of this energy is in the slow-heating phase. About two-fifths of the total energy used to heat the corona is radiated by the coronal plasmas, and the other three fifth transported to the lower atmosphere by thermal conduction.

  13. Ostwald ripening in two-phase mixtures

    International Nuclear Information System (INIS)

    Voorhees, P.W.

    1982-01-01

    Experimental measurements of the temperature of a rapidly solidified solid-liquid mixture have been made over a range of volume fractions solid 0.23 to 0.95. These experiments demonstrate the viability of measuring the change in interfacial curvature with time via precision thermometry. The experimental measurements also indicate that there is no radical change in interface morphology over a wide range of volume fractions solid. A solution to the multi-particle diffusion problem (MDP) has been constructed through the use of potential theory. The solution to the MDP was used to describe the diffusion field within a coarsening two-phase mixture consisting of dispersed spherical second-phase particles. Since this theory is based upon the MDP, interparticle diffusional interactions are specifically included in the treatment. As a result, the theory yields, for the first time, insights into the influence of the local distribution of curvature on a particle's coarsening rate. The effect of interparticle interactions on the collective behavior of an ensemble of coarsening particles was also investigated. It was found that any arbitrary distribution of particle radii will tend to a specific time independent distribution when the particle radii are scaled by the average particle radius. Furthermore, it was determined that with increasing volume fraction of coarsening phase, these time independent distributions become broader and more symmetric. It was also found that the ripening kinetics, as measured by the growth rate of the average particle size, increases by a factor of five upon increasing the volume fraction of coarsening phase from zero to 0.5

  14. Multigrid preconditioning of the generator two-phase mixture balance equations in the Genepi software

    International Nuclear Information System (INIS)

    Belliard, M.; Grandotto, M.

    2003-01-01

    In the framework of the two-phase fluid simulations of the steam generators of pressurized water nuclear reactors, we present in this paper a geometric version of a pseudo-Full MultiGrid (pseudo- FMG) Full Approximation Storage (FAS) preconditioning of balance equations in the GENEPI code. In our application, the 3D steady state flow is reached by a transient computation using a semi-implicit fractional step algorithm for the averaged two-phase mixture balance equations (mass, momentum and energy for the secondary flow). Our application, running on workstation clusters, is based on a CEA code-linker and the PVM package. The difficulties to apply the geometric FAS multigrid method to the momentum and mass balance equations are addressed. The use of a sequential pseudo-FMG FAS twogrid method for both energy and mass/momentum balance equations, using dynamic multigrid cycles, leads to perceptibly improvements in the computation convergences. An original parallel red-black pseudo-FMG FAS three-grid algorithm is presented too. The numerical tests (steam generator mockup simulations) underline the sizable increase in speed of convergence of the computations, essentially for the ones involving a large number of freedom degrees (about 100 thousand cells). The two-phase mixture balance equation residuals are quickly reduced: the reached speed-up stands between 2 and 3 following the number of grids. The effects on the convergence behavior of the numerical parameters are investigated

  15. Study on hydrodynamic crisis of two-phase flow

    International Nuclear Information System (INIS)

    Nigmatulin, B.I.; Ivandaev, A.I.

    1977-01-01

    The phenomenon of hydrodynamic crisis (locking) of a two-phase flow is investigated. A model of a disperseannular flow with an effective monodisperse nucleus is used for describing the motion of a mixture under near-critical conditions. Main differential equations of a flow in a channel are given; in particular, the differential laws of variation of the effective diameters of drops in the nucleus as a result of mass exchange between the mixture components are singled out. Questions of concretization of the model are discussed. The conditions for the attainment of the maximum rate of flow of the gas through the channel are studied, as well as the effect of the flow prehistory on the formation of critical conditions in the outlet cross-section

  16. Study on hydrodynamic crisis of two-phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Nigmatulin, B I; Ivandaev, A I [Moskovskij Gosudarstvennyj Univ. (USSR). Nauchno-Issledovatel' skij Inst. Mekhaniki

    1977-01-01

    The phenomenon of hydrodynamic crisis (locking) of a two-phase flow is investigated. A model of a disperse annular flow with an effective monodisperse nucleus is used for describing the motion of a mixture under near-critical conditions. Main differential equations of a flow in a channel are given; in particular, the differential laws of variation of the effective diameters of drops in the nucleus as a result of mass exchange between the mixture components are singled out. Questions of concretization of the model are discussed. The conditions for the attainment of the maximum rate of flow of the gas through the channel are studied, as well as the effect of the flow prehistory on the formation of critical conditions in the outlet cross-section.

  17. Pressure drop, steam content and turbulent cross exchange in water/steam flows

    International Nuclear Information System (INIS)

    Teichel, H.

    1978-01-01

    For describing the behaviour of two-phase flows of water and steam with the help of calculating patterns, a number of empirical correlations are required. - In this article, correlations for the friction pressure drop in water/steam flows are compared, as well as for the steam mass and the volumetric steam content with each other and with the test results on simple geometries. As the mutual effect between cooling chanels plays an important part at the longitudinal flow through bar bundles, the appertaining equations are evaluated, in addition. (orig.) 891 HP [de

  18. Vapor Compressor Driven Hybrid Two-Phase Loop, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This Small Business Innovation Research Phase I project will demonstrate a vapor compressor driven hybrid two-phase loop technology. The hybrid two-phase loop...

  19. Reduced order modeling of flashing two-phase jets

    Energy Technology Data Exchange (ETDEWEB)

    Gurecky, William, E-mail: william.gurecky@utexas.edu; Schneider, Erich, E-mail: eschneider@mail.utexas.edu; Ballew, Davis, E-mail: davisballew@utexas.edu

    2015-12-01

    Highlights: • Accident simulation requires ability to quickly predict two-phase flashing jet's damage potential. • A reduced order modeling methodology informed by experimental or computational data is described. • Zone of influence volumes are calculated for jets of various upstream thermodynamic conditions. - Abstract: In the event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor, the escaping coolant produces a highly energetic flashing jet with the potential to damage surrounding structures. In LOCA analysis, the goal is often to evaluate many break scenarios in a Monte Carlo style simulation to evaluate the resilience of a reactor design. Therefore, in order to quickly predict the damage potential of flashing jets, it is of interest to develop a reduced order model that relates the damage potential of a jet to the pressure and temperature upstream of the break and the distance from the break to a given object upon which the jet is impinging. This work presents framework for producing a Reduced Order Model (ROM) that may be informed by measured data, Computational Fluid Dynamics (CFD) simulations, or a combination of both. The model is constructed by performing regression analysis on the pressure field data, allowing the impingement pressure to be quickly reconstructed for any given upstream thermodynamic condition within the range of input data. The model is applicable to both free and fully impinging two-phase flashing jets.

  20. Durability of bends in high-temperature steam lines under the conditions of long-term operation

    Science.gov (United States)

    Katanakha, N. A.; Semenov, A. S.; Getsov, L. B.

    2015-04-01

    The article presents the results of stress-strain state computations and durability of bent and steeply curved branches of high-temperature steam lines carried out on the basis of the finite element method using the modified Soderberg formula for describing unsteady creep processes with taking the accumulation of damage into account. The computations were carried out for bends made of steel grades that are most widely used for manufacturing steam lines (12Kh1MF, 15Kh1M1F, and 10Kh9MFB) and operating at different levels of inner pressure and temperature. The solutions obtained using the developed creep model are compared with those obtained using the models widely used in practice.

  1. Semi-mechanistic Model Applied to the Search for Economically Optimal Conditions and Blending of Gasoline Feedstock for Steam-cracking Process

    Directory of Open Access Journals (Sweden)

    Karaba Adam

    2016-01-01

    Full Text Available Steam-cracking is energetically intensive large-scaled process which transforms a wide range of hydrocarbons feedstock to petrochemical products. The dependence of products yields on feedstock composition and reaction conditions has been successfully described by mathematical models which are very useful tools for the optimization of cracker operation. Remaining problem is to formulate objective function for such an optimization. Quantitative criterion based on the process economy is proposed in this paper. Previously developed and verified industrial steam-cracking semi-mechanistic model is utilized as supporting tool for economic evaluation of selected gasoline feedstock. Economic criterion is established as the difference between value of products obtained by cracking of studied feedstock under given conditions and the value of products obtained by cracking of reference feedstock under reference conditions. As an example of method utilization, optimal reaction conditions were searched for each of selected feedstock. Potential benefit of individual cracking and cracking of grouped feedstocks in the contrast to cracking under the middle of optimums is evaluated and also compared to cracking under usual conditions.

  2. Thermo-fluid dynamics of two-phase flow

    CERN Document Server

    Ishii, Mamoru; Ishii, Mamoru; Ishii, M

    2006-01-01

    Provides a very systematic treatment of two phase flow problems from a theoretical perspectiveProvides an easy to follow treatment of modeling and code devlopemnt of two phase flow related phenomenaCovers new results of two phase flow research such as coverage of fuel cells technology.

  3. Thermodynamic performance simulation and concise formulas for triple-pressure reheat HRSG of gas–steam combined cycle under off-design condition

    International Nuclear Information System (INIS)

    Zhang, Guoqiang; Zheng, Jiongzhi; Yang, Yongping; Liu, Wenyi

    2016-01-01

    Highlights: • An off-design performance simulation of triple-pressure reheat HRSG is executed. • The bottoming cycle characteristics of energy transfer/conversion are analyzed. • Concise formulas for the off-design performance of bottoming cycle are proposed. • The accuracy of the formulas is verified under different load control strategies. • The errors of the formulas are generally within 1% at a load of 100–50%. - Abstract: Concise semi-theoretical, semi-empirical formulas are developed in this study to predict the off-design performance of the bottoming cycle of the gas–steam turbine combined cycle. The formulas merely refer to the key thermodynamic design parameters (full load parameters) of the bottoming cycle and off-design gas turbine exhaust temperature and flow, which are convenient in determining the overall performance of the bottoming cycle. First, a triple-pressure reheat heat recovery steam generator (HRSG) is modeled, and thermodynamic analysis is performed. Second, concise semi-theoretical, semi-empirical performance prediction formulas for the bottoming cycle are proposed through a comprehensive analysis of the heat transfer characteristics of the HRSG and the energy conversion characteristics of the steam turbine under the off-design condition. The concise formulas are found to be effective, i.e., fast, simple, and precise in obtaining the thermodynamic parameters for bottoming cycle efficiency, HRSG heat transfer capacity, HRSG efficiency, steam turbine power output, and steam turbine efficiency under the off-design condition. Accuracy is verified by comparing the concise formulas’ calculation results with the simulation results and practical operation data under different load control strategies. The calculation errors are within 1.5% (mainly less than 1% for both simulation and actual operation data) under combined cycle load (gas turbine load) ranging from 50% to 100%. However, accuracy declines sharply when the turbine

  4. What is geothermal steam worth?

    International Nuclear Information System (INIS)

    Thorhallsson, S.; Ragnarsson, A.

    1992-01-01

    Geothermal steam is obtained from high-temperature boreholes, either directly from the reservoir or by flashing. The value of geothermal steam is similar to that of steam produced in boilers and lies in its ability to do work in heat engines such as turbines and to supply heat for a wide range of uses. In isolated cases the steam can be used as a source of chemicals, for example the production of carbon dioxide. Once the saturated steam has been separated from the water, it can be transported without further treatment to the end user. There are several constraints on its use set by the temperature of the reservoir and the chemical composition of the reservoir fluid. These constraints are described (temperature of steam, scaling in water phase, gas content of steam, well output) as are the methods that have been adopted to utilize this source of energy successfully. Steam can only be transported over relatively short distances (a few km) and thus has to be used close to the source. Examples are given of the pressure drop and sizing of steam mains for pipelines. The path of the steam from the reservoir to the end user is traced and typical cost figures given for each part of the system. The production cost of geothermal steam is estimated and its sensitivity to site-specific conditions discussed. Optimum energy recovery and efficiency is important as is optimizing costs. The paper will treat the steam supply system as a whole, from the reservoir to the end user, and give examples of how the site-specific conditions and system design have an influence on what geothermal steam is worth from the technical and economic points of view

  5. Homogeneous non-equilibrium two-phase critical flow model

    International Nuclear Information System (INIS)

    Schroeder, J.J.; Vuxuan, N.

    1987-01-01

    An important aspect of nuclear and chemical reactor safety is the ability to predict the maximum or critical mass flow rate from a break or leak in a pipe system. At the beginning of such a blowdown, if the stagnation condition of the fluid is subcooled or slightly saturated thermodynamic non-equilibrium exists in the downstream, e.g. the fluid becomes superheated to a degree determined by the liquid pressure. A simplified non-equilibrium model, explained in this report, is valid for rapidly decreasing pressure along the flow path. It presumes that fluid has to be superheated by an amount governed by physical principles before it starts to flash into steam. The flow is assumed to be homogeneous, i.e. the steam and liquid velocities are equal. An adiabatic flow calculation mode (Fanno lines) is employed to evaluate the critical flow rate for long pipes. The model is found to satisfactorily describe critical flow tests. Good agreement is obtained with the large scale Marviken tests as well as with small scale experiments. (orig.)

  6. Two-phase flow instabilities in a vertical annular channel

    Energy Technology Data Exchange (ETDEWEB)

    Babelli, I.; Nair, S.; Ishii, M. [Purdue Univ., West Lafayette, IN (United States)

    1995-09-01

    An experimental test facility was built to study two-phase flow instabilities in vertical annular channel with emphasis on downward flow under low pressure and low flow conditions. The specific geometry of the test section is similar to the fuel-target sub-channel of the Savannah River Site (SRS) Mark 22 fuel assembly. Critical Heat Flux (CHF) was observed following flow excursion and flow reversal in the test section. Density wave instability was not recorded in this series of experimental runs. The results of this experimental study show that flow excursion is the dominant instability mode under low flow, low pressure, and down flow conditions. The onset of instability data are plotted on the subcooling-Zuber (phase change) numbers stability plane.

  7. Void fraction and interfacial velocity in gas-liquid upward two-phase flow across tube bundles

    International Nuclear Information System (INIS)

    Ueno, T.; Tomomatsu, K.; Takamatsu, H.; Nishikawa, H.

    1997-01-01

    Tube failures due to flow-induced vibration are a major problem in heat exchangers and many studies on the problem of such vibration have been carried out so far. Most studies however, have not focused on two-phase flow behavior in tube bundles, but have concentrated mainly on tube vibration behavior like fluid damping, fluid elastic instability and so on. Such studies are not satisfactory for understanding the design of heat exchangers. Tube vibration behavior is very complicated, especially in the case of gas-liquid two-phase flow, so it is necessary to investigate two-phase flow behavior as well as vibration behavior before designing heat exchangers. This paper outlines the main parameters that characterize two-phase behavior, such as void fraction and interfacial velocity. The two-phase flow analyzed here is gas-liquid upward flow across a horizontal tube bundle. The fluids tested were HCFC-123 and steam-water. HCFC-123 stands for Hydrochlorofluorocarbon. Its chemical formula is CHCl 2 CF 3 , which has liquid and gas densities of 1335 and 23.9 kg/m 3 at a pressure of 0.40 MPa and 1252 and 45.7 kg/m 3 at a pressure of 0.76 MPa. The same model tube bundle was used in the two tests covered in this paper, to examine the similarity law of two-phase flow behavior in tube bundles using HCFC-123 and steam-water two-phase flow. We also show numerical simulation results for the two fluid models in this paper. We do not deal with vibration behavior and the relationship between vibration behavior and two-phase flow behavior. (author)

  8. Modeling of two-phase flow with thermal and mechanical non-equilibrium

    International Nuclear Information System (INIS)

    Houdayer, G.; Pinet, B.; Le Coq, G.; Reocreux, M.; Rousseau, J.C.

    1977-01-01

    To improve two-phase flow modeling by taking into account thermal and mechanical non-equilibrium a joint effort on analytical experiment and physical modeling has been undertaken. A model describing thermal non-equilibrium effects is first presented. A correlation of mass transfer has been developed using steam water critical flow tests. This model has been used to predict in a satisfactory manner blowdown tests. It has been incorporated in CLYSTERE system code. To take into account mechanical non-equilibrium, a six equations model is written. To get information on the momentum transfers special nitrogen-water tests have been undertaken. The first results of these studies are presented

  9. Mechanism of falling water limitation in two-phase counter flow through single hole vertical channel

    International Nuclear Information System (INIS)

    Sudo, Yukio; Ohnuki, Akira

    1983-01-01

    In the safety evaluation at the time of loss coolant accident, which is a credible accident in LWRs, recently main effort has been concentrated to the optimum evaluation calculation, and the grasp of vapor-liquid two-phase flow phenomena has become important. As one of the important phenomena, there is the limitation of falling water in two-phase counter flow through a vertical channel. This phenomenon is divided into the limitation of falling water stored in an upper plenum to a core through an upper core-supporting plate and a tie plate at the time of reflooding, and the limitation of falling emergency core-cooling water in downcomer channels at the time of reflooding in PWRs, under the presence of rising steam flow. In both cases, the evaluation of the quantity of falling water is important, because it contributes directly to core cooling. In this research, in order to clarify the mechanism of limitation of falling water in two-phase vertical counter flow, first, two-phase flow of air-water system through a single-hole vertical channel was taken up, and the effect of main parameters was experimentally studied. At the same time, the theoretical investigation was performed, and the comparison with the experimental results obtained so far was carried out. The different mechanisms for short and long channels gave the good results. (Kako, I.)

  10. On-Line Condition Monitoring System for High Level Trip Water in Steam Boiler’s Drum

    Directory of Open Access Journals (Sweden)

    Ismail Alnaimi Firas B.

    2014-07-01

    Full Text Available This paper presents a monitoring technique using Artificial Neural Networks (ANN with four different training algorithms for high level water in steam boiler’s drum. Four Back-Propagations neural networks multidimensional minimization algorithms have been utilized. Real time data were recorded from power plant located in Malaysia. The developed relevant variables were selected based on a combination of theory, experience and execution phases of the model. The Root Mean Square (RMS Error has been used to compare the results of one and two hidden layer (1HL, (2HL ANN structures

  11. Studies on turbulence structure and liquid film behavior in annular two-phase flow flowing in a throat section

    International Nuclear Information System (INIS)

    Yoshida, Kenji; Miyabe, Masaya; Matsumoto, Tadayoshi; Kataoka, Isao; Ohmori, Shuichi; Mori, Michitsugu

    2004-01-01

    Experimental studies on turbulence structure and liquid film behavior in annular two-phase flow were carried out concerned with the steam injector systems for a next-generation nuclear reactor. In the steam injector, steam/water annular two-phase flow is formed at the mixing nozzle. To make an appropriate design for high-performance steam injector system, it is very important to accumulate the fundamental data of thermo-hydro dynamic characteristics of annular flow in the steam injector. Especially, the turbulence modification in multi-phase flow due to the phase interaction is one of the most important phenomena and has attracted research attention. In this study, the liquid film behavior and the resultant turbulence modification due to the phase interaction were investigated. The behavior of the interfacial waves on liquid film flow such as the ripple or disturbance waves were observed to make clear the interfacial velocity and the special structure of the interfacial waves by using the high-speed video camera and the digital camera. The measurements for gas-phase velocity profiles and turbulent intensity in annular flow passing through the throat section were precisely performed to investigate quantitatively the turbulent modification in annular flow by using the constant temperature hot-wire anemometer. The measurements for liquid film thickness by the electrode needle method were also carried out. (author)

  12. Two-phase flow patterns and their relationship to two-phase heat transfer

    International Nuclear Information System (INIS)

    Hewitt, G.F.

    1977-01-01

    The objective of this lecture was to discuss the general nature of two phase flows, to define the various regimes of flow and to discuss the influence of these regimes on the heat transfer processes taking place. The methods of regime delineation are briefly described and regime descriptions introduced for both vertical and horizontal flows in tubes. ''Flow regime maps'' have been widely used as an aid to determination of the regime which occurs in a given situation. Some of the more widely used maps are described and the limitations of this approach discussed. There have been many attempts to obtain a better phenomenological description of two phase flow patterns. In this lecture, these attempts will be reviewed in the context of the bubble/plug, plug/churn and churn/annular flow transitions in vertical flow. The latter two transitions are related to the flooding/flow reversal phenomena. For horizontal flows, recent work on the onset of slugging will be reviewed. In flows with evaporation or condensation, the situation is influenced by departures from thermodynamic equilibrium and the types of departure observed are discuss briefly. Flow patterns and their relationships with heat transfer regimes are then reviewed for the case of condensation in horizontal tubes and evaporation in vertical tubes

  13. Development of an ex-vessel corium debris bed with two-phase natural convection in a flooded cavity

    International Nuclear Information System (INIS)

    Kim, Eunho; Lee, Mooneon; Park, Hyun Sun; Moriyama, Kiyofumi; Park, Jin Ho

    2016-01-01

    Highlights: • For ex-vessel severe accidents in LWRs with wet-cavity strategy, development of debris bed with two-phase natural convection flow due to thermal characteristics of prototypic corium particles was investigated experimentally by using simulant particles and local air bubble control system. • Based on the experimental results of this study, an analytical model was established to describe the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. • This model was then used to analyze the formation of debris beds at the reactor scale, and a sensitivity analysis was carried out based on key accident parameters. - Abstract: During severe accidents of light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue and a threat to the integrity of containment. With a flooded cavity, a porous debris bed is expected to develop on the bottom of the pool due to breakup and fragmentation of the melt jet. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) experimental apparatus facility was constructed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed. Using this system, five kilograms of stainless steel simulant debris were injected from the top of the water level, while air bubbles simulating the vapor flow were injected from the bottom of the particle catcher plate. The airflow rate was determined based on the quantity of settled debris, which will form a heat source due to the decay of corium. The radial distribution of the settled debris was examined using a ‘gap–tooth’ approach. Based on the experimental results of this study, an analytical model was established to

  14. Development of an ex-vessel corium debris bed with two-phase natural convection in a flooded cavity

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eunho; Lee, Mooneon; Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr; Moriyama, Kiyofumi; Park, Jin Ho

    2016-03-15

    Highlights: • For ex-vessel severe accidents in LWRs with wet-cavity strategy, development of debris bed with two-phase natural convection flow due to thermal characteristics of prototypic corium particles was investigated experimentally by using simulant particles and local air bubble control system. • Based on the experimental results of this study, an analytical model was established to describe the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. • This model was then used to analyze the formation of debris beds at the reactor scale, and a sensitivity analysis was carried out based on key accident parameters. - Abstract: During severe accidents of light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue and a threat to the integrity of containment. With a flooded cavity, a porous debris bed is expected to develop on the bottom of the pool due to breakup and fragmentation of the melt jet. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) experimental apparatus facility was constructed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed. Using this system, five kilograms of stainless steel simulant debris were injected from the top of the water level, while air bubbles simulating the vapor flow were injected from the bottom of the particle catcher plate. The airflow rate was determined based on the quantity of settled debris, which will form a heat source due to the decay of corium. The radial distribution of the settled debris was examined using a ‘gap–tooth’ approach. Based on the experimental results of this study, an analytical model was established to

  15. The use of advanced scale conditioning agents for maintenance of the secondary side of nuclear plant steam generators

    International Nuclear Information System (INIS)

    Battaglia, P.J.; Rogosky, D.L.

    2006-01-01

    Maintaining the secondary side of steam generators within a pressurized water reactor (PWR) free of deposited corrosion products and corrosion-inducing contaminants is key to ensuring their long-term operation. New cleaning processes have become available to aid nuclear plant personnel in optimizing secondary side maintenance strategies. These strategies include both maintaining nuclear steam generators corrosion free while maintaining full power operation. The conference presentation will discuss ASCA use and the major field experience acquired in the last several years in the United States and in Japan. Hokkaido Electric, Dominion Engineering, Inc. and Westinghouse cosponsored the development of ASCAs for use in the Nuclear Utility industry, and all three are active in field use programs. Westinghouse owns the worldwide rights for ASCA implementation except in Japan where MHI and NEL have been granted licenses to apply ASCAs. Dominion Engineering Inc., owns the ASCA patents and performs the laboratory qualification testing associated with the ASCA programs, and Hokkaido Electric are joint patent holders for ASCAs and have been implementing their use at the Tomari plants for cleaning and thermal hydraulic performance enhancements. The specific experience discussed in the presentation will include: 1. Full Bundle Maintenance ASCAs at Vogtle Units 2 and 2 and Wolf Creek (USA). 2. Top of the Tubesheet ASCAs with high pressure sludge lancing at Wolf Creek and UEC at Vogtle Units 1 and 2 (USA). 3. Thermal Hydraulic Recovery and Maintenance ASCAs at the Hokkaido Electric Tomari Units 1 and 2 (Japan). (author)

  16. Developmental assessment of RELAP5/MOD3 code against ROSA-IV/TPTF horizontal two-phase flow experiments

    International Nuclear Information System (INIS)

    Kukita, Yutaka; Asaka, Hideaki; Anoda, Yoshinari; Ishiguro, Misako; Tasaka, Kanji; Mimura, Yuichi; Nemoto, Toshiyuki.

    1990-03-01

    A developmental version of the RELAP5/Mod3 code (as of June 1989) was assessed for accuracy using experimental data taken for high-pressure (7MPa) steam-water two-phase flow in a large-diameter (0.18 m) horizontal-pipe test section of the ROSA-IV Two-Phase Flow Test Facility (TPTF). The agreement between the measured and calculated test section void fractions was much better than that for the previous generation of RELAP5 (MOD2). The improvement was achieved primarily due to the code changes with respect to the flow stratification criterion and interfacial-drag calculation scheme. (author)

  17. Mathematical model of two-phase flow in accelerator channel

    Directory of Open Access Journals (Sweden)

    О.Ф. Нікулін

    2010-01-01

    Full Text Available  The problem of  two-phase flow composed of energy-carrier phase (Newtonian liquid and solid fine-dispersed phase (particles in counter jet mill accelerator channel is considered. The mathematical model bases goes on the supposition that the phases interact with each other like independent substances by means of aerodynamics’ forces in conditions of adiabatic flow. The mathematical model in the form of system of differential equations of order 11 is represented. Derivations of equations by base physical principles for cross-section-averaged quantity are produced. The mathematical model can be used for estimation of any kinematic and thermodynamic flow characteristics for purposely parameters optimization problem solving and transfer functions determination, that take place in  counter jet mill accelerator channel design.

  18. Two-phase flow experiments through intergranular stress corrosion cracks

    International Nuclear Information System (INIS)

    Collier, R.P.; Norris, D.M.

    1984-01-01

    Experimental studies of critical two-phase water flow, through simulated and actual intergranular stress corrosion cracks, were performed to obtain data to evaluate a leak flow rate model and investigate acoustic transducer effectiveness in detecting and sizing leaks. The experimental program included a parametric study of the effects of crack geometry, fluid stagnation pressure and temperature, and crack surface roughness on leak flow rate. In addition, leak detection, location, and leak size estimation capabilities of several different acoustic transducers were evaluated as functions of leak rate and transducer position. This paper presents flow rate data for several different cracks and fluid conditions. It also presents the minimum flows rate detected with the acoustic sensors and a relationship between acoustic signal strength and leak flow rate

  19. Flooding in counter-current two-phase flow

    International Nuclear Information System (INIS)

    Ragland, W.A.; Ganic, E.N.

    1982-01-01

    Flooding is a phenomenon which is best described as the transition from counter-current to co-current flow. Early notice was taken of this phenomenon in the chemical engineering industry. Flooding also plays an important role in the field of two-phase heat transfer since it is a limit for many systems involving counter-current flow. Practical applications of flooding limited processes include wickless thermosyphons and the emergency core cooling system (ECCS) of pressurized water nuclear reactors. The phenomenon of flooding also is involved in the behavior of nuclear reactor core materials during severe accident conditions where flooding is one of the mechanisms governing the motion of the molten fuel pin cladding

  20. Two-phase flow and heat transfer under low gravity

    Science.gov (United States)

    Frost, W.

    1981-01-01

    Spacelab experiment to investigate two-phase flow patterns under gravity uses a water-air mixture experiment. Air and water are circulated through the system. The quality or the mixture or air-water is controlled. Photographs of the test section are made and at the same time pressure drop across the test section is measured. The data establishes a flow regime map under reduced gravity conditions with corresponding pressure drop correlations. The test section is also equipped with an electrical resistance heater in order to allow a flow boiling experiment to be carried out using Freon II. High-speed photographs of the test section are used to determine flow patterns. The temperature gradient and pressure drop along the duct can be measured. Thus, quality change can be measured, and heat transfer calculated.

  1. Characteristics of two-phase flows in large diameter channels

    Energy Technology Data Exchange (ETDEWEB)

    Schlegel, J.P., E-mail: schlegelj@mst.edu [Department of Mining and Nuclear Engineering, Missouri University of Science and Technology, 301 W 14th St., Rolla, MO 65401 (United States); Hibiki, T.; Ishii, M. [School of Nuclear Engineering, Purdue University, 400 Central Dr., West Lafayette, IN 47907 (United States)

    2016-12-15

    Two-phase flows in large diameter channels have a great deal of importance in a wide variety of industrial applications. Nuclear systems, petroleum refineries, and chemical processes make extensive use of larger systems. Flows in such channels have very different properties from flows in smaller channels which are typically used in experimental research. In this paper, the various differences between flows in large and small channels are highlighted using the results of previous experimental and analytical research. This review is followed by a review of recent experiments in and model development for flows in large diameter channels performed by the authors. The topics of these research efforts range from void fraction and interfacial area concentration measurement to flow regime identification and modeling, drift-flux modeling for high void fraction conditions, and evaluation of interfacial area transport models for large diameter channels.

  2. Flooding in counter-current two-phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Ragland, W.A.; Ganic, E.N.

    1982-01-01

    Flooding is a phenomenon which is best described as the transition from counter-current to co-current flow. Early notice was taken of this phenomenon in the chemical engineering industry. Flooding also plays an important role in the field of two-phase heat transfer since it is a limit for many systems involving counter-current flow. Practical applications of flooding limited processes include wickless thermosyphons and the emergency core cooling system (ECCS) of pressurized water nuclear reactors. The phenomenon of flooding also is involved in the behavior of nuclear reactor core materials during severe accident conditions where flooding is one of the mechanisms governing the motion of the molten fuel pin cladding.

  3. A Rotational Pressure-Correction Scheme for Incompressible Two-Phase Flows with Open Boundaries

    Science.gov (United States)

    Dong, S.; Wang, X.

    2016-01-01

    Two-phase outflows refer to situations where the interface formed between two immiscible incompressible fluids passes through open portions of the domain boundary. We present several new forms of open boundary conditions for two-phase outflow simulations within the phase field framework, as well as a rotational pressure correction based algorithm for numerically treating these open boundary conditions. Our algorithm gives rise to linear algebraic systems for the velocity and the pressure that involve only constant and time-independent coefficient matrices after discretization, despite the variable density and variable viscosity of the two-phase mixture. By comparing simulation results with theory and the experimental data, we show that the method produces physically accurate results. We also present numerical experiments to demonstrate the long-term stability of the method in situations where large density contrast, large viscosity contrast, and backflows occur at the two-phase open boundaries. PMID:27163909

  4. Flashing liquid jets and two-phase droplet dispersion

    International Nuclear Information System (INIS)

    Cleary, Vincent; Bowen, Phil; Witlox, Henk

    2007-01-01

    The large-scale release of a liquid contained at upstream conditions above its local atmospheric boiling point is a scenario often given consideration in process industry risk analysis. Current-hazard quantification software often employs simplistic equilibrium two-phase approaches. Scaled water experiments have been carried out measuring droplet velocity and droplet size distributions for a range of exit orifice aspect ratios (L/d) and conditions representing low to high superheat. 2D Phase-Doppler Anemometry has been utilised to characterise droplet kinematics and spray quality. Droplet size correlations have been developed for non-flashing, the transition between non-flashing and flashing, and fully flashing jets. Using high-speed shadowography, transition between regimes is defined in terms of criteria identified in the external flow structure. An overview companion paper provides a wider overview of the problem and reports implementation of these correlations into consequence models and subsequent validation. The fluid utilised throughout is water, hence droplet correlations are developed in non-dimensional form to allow extrapolation to other fluids through similarity scaling, although verification of model performance for other fluids is required in future studies. Data is reduced via non-dimensionalisation in terms of the Weber number and Jakob number, essentially representing the fluid mechanics and thermodynamics of the system, respectively. A droplet-size distribution correlation has also been developed, conveniently presented as a volume undersize distribution based on the Rosin-Rammler distribution. Separate correlations are provided for sub-cooled mechanical break-up and fully flashing jets. This form of correlation facilitates rapid estimates of likely mass rainout quantities, as well as full distribution information for more rigorous two-phase thermodynamic modelling in the future

  5. Performance of iron–chromium–aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Weicheng; Mouche, Peter A.; Han, Xiaochun [University of Illinois, Department of Nuclear, Radiological, and Plasma Engineering, Urbana, IL 61801 (United States); Heuser, Brent J., E-mail: bheuser@illinois.edu [University of Illinois, Department of Nuclear, Radiological, and Plasma Engineering, Urbana, IL 61801 (United States); Mandapaka, Kiran K.; Was, Gary S. [University of Michigan, Department of Nuclear Engineering and Radiological Sciences, Ann Arbor, MI 48109 (United States)

    2016-03-15

    Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe–Zr is addressed with the FeCrAl-YSZ system. - Graphical abstract: Weight gain normalized to total sample surface area versus time during 700 °C steam exposure for FeCrAl samples with different composition (A) and Fe/Cr/Al:62/4/34 (B). In both cases, the responses of uncoated Zry2 (Zry2-13A and Zry2-19A) are shown for comparison. This uncoated Zry2 response shows the expected pre-transition quasi-cubic kinetic behavior and eventual breakaway (linear) kinetics. Highlights: • FeCrAl coatings deposited on Zy2 have been tested with respect to oxidation in high-temperature steam. • FeCrAl compositions promoting alumina formation inhibited oxidation of Zy2 and delay weight gain. • Autoclave testing to 20 days of coated Zy2 in a simulated BWR environment demonstrates minimal weight gain and no film degradation. • The 900 °C eutectic in binary Fe-Zr is addressed with the FeCrAl-YSZ system.

  6. Steam-generator simulation with non-equilibrium two-phase flow models

    Energy Technology Data Exchange (ETDEWEB)

    Jena, A K; Singh, O P

    1986-01-01

    The angular dependence of neutron-noise transmission through shields in LMFBRs is considered. The results are presented in particular, on the angle dependence of the APSD of the angular neutron-noise flux. It is observed that the larger the angle of the neutron-noise flux, the greater the break frequency of the corresponding APSD.

  7. Study of two-phase foam flow

    Energy Technology Data Exchange (ETDEWEB)

    Gurbanov, R S; Guliev, B B; Mekhtiev, K G; Kerimov, R G

    1970-01-01

    The objectives of this study were to determine characteristics of aqueous foam flow through porous media and to estimate the depth of foam penetration into a formation. Foam was generated by mixing air and 1% solution of surfactant PO-1. Foam density was maintained at 0.14 g/cc in all experiments. The foam was passed through sand columns (800 mm long x 30 mm diam) of permeabilities 26, 39, 80, 111, and 133 darcys. Flow rates were measured at various pressure drops and the relationship between system parameters was expressed analytically and graphically. From the data, distance of foam penetration into a formation as a function of pressure drop and permeability was calculated. The data indicate that under most conditions, foam will penetrate the formation to a negligible distance. This study indicates that when foam is used to remove sand from a well, a negligible loss of foam to the formation occurs.

  8. A review of damping of two-phase flows

    International Nuclear Information System (INIS)

    Hara, Fumio

    1993-01-01

    Damping of two-phase flows has been recognized as one of the most unknown parameters in analyzing vibrational characteristics of structures subjected to two-phase flows since it seems to be influenced by many physical parameters involved in the physics of dynamic energy dissipation of a vibrating structure, for example, liquid viscosity, surface tension, flow velocity, mass ratio, frequency, void fraction, flow regime and so forth. This paper deals with a review of scientific works done to date on the damping of two phase flows and discussions about what has been clarified and what has not been known to us, or what kinds of research are needed about two-phase flow damping. The emphasis is put on the definition of two-phase fluid damping, damping measurement techniques, damping characteristics in relation to two phase flow configurations, and damping generation mechanisms

  9. Forced two phase helium cooling of large superconducting magnets

    International Nuclear Information System (INIS)

    Green, M.A.; Burns, W.A.; Taylor, J.D.

    1979-08-01

    A major problem shared by all large superconducting magnets is the cryogenic cooling system. Most large magnets are cooled by some variation of the helium bath. Helium bath cooling becomes more and more troublesome as the size of the magnet grows and as geometric constraints come into play. An alternative approach to cooling large magnet systems is the forced flow, two phase helium system. The advantages of two phase cooling in many magnet systems are shown. The design of a two phase helium system, with its control dewar, is presented. The paper discusses pressure drop of a two phase system, stability of a two phase system and the method of cool down of a two phase system. The results of experimental measurements at LBL are discussed. Included are the results of cool down and operation of superconducting solenoids

  10. Numerical method for two-phase flow discontinuity propagation calculation

    International Nuclear Information System (INIS)

    Toumi, I.; Raymond, P.

    1989-01-01

    In this paper, we present a class of numerical shock-capturing schemes for hyperbolic systems of conservation laws modelling two-phase flow. First, we solve the Riemann problem for a two-phase flow with unequal velocities. Then, we construct two approximate Riemann solvers: an one intermediate-state Riemann solver and a generalized Roe's approximate Riemann solver. We give some numerical results for one-dimensional shock-tube problems and for a standard two-phase flow heat addition problem involving two-phase flow instabilities

  11. 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

  12. Definition of a facility for experimental studies of two-phase flows and heat transfer in porous materials

    International Nuclear Information System (INIS)

    Reda, D.C.; Eaton, R.R.

    1981-01-01

    A facility-development effort is currently underway at Sandia National Laboratories in order to create an experimental capability for the study of two-phase, steam/water flows through a variety of porous media. The facility definition phase of this project is described. Equations are derived for the steady, adiabatic, macroscopically-linear two-phase flow of a single-component fluid through a porous medium, including energy transfer both by convection and conduction. These equations are then solved to give relative permeabilities for the steam and water phases as functions of known and/or measurable quantities. A viable experimental approach was thereby formulated, leading to the definition of facility components and instrumentation requirements, including the application of gamma-beam densitometry for the measurement of liquid-saturation distributions in porous media. Finally, a state-of-the-art computer code was utilized to numerically simulate the proposed experiments, providing an estimate of the facility operating envelope

  13. A model of single and two-phase flow (critical or not) through cracks

    International Nuclear Information System (INIS)

    Seynhaeve, J.M.; Giot, M.; Granger, S.; Pages, D.

    1995-07-01

    The leaks through steam-generator cracks are the subject of research carried out in cooperation between EDF and UCL. A model to predict the mass flow rate with inlet subcooling has been developed, validated and published. The model takes into account the persistence of some metastable liquid in the crack. The present paper improves and extends the model, by making it applicable to all kinds of conditions prevailing in the S.G. tubes: not only subcooled water, but also saturated water, steam-water mixtures, saturated dry steam or superheated steam. Therefore, the flow at the crack inlet is analyzed and appropriate methods to initialize the numerical integration of the flow equations along the crack are proposed. The extensions of the model are still in the process of validation. However, a sensitivity analysis of its results has been made and is presented. (author)

  14. Evaluation method for two-phase flow and heat transfer in a feed-water heater

    International Nuclear Information System (INIS)

    Takamori, Kazuhide; Minato, Akihiko

    1993-01-01

    A multidimensional analysis code for two-phase flow using a two-fluid model was improved by taking into consideration the condensation heat transfer, film thickness, and film velocity, in order to develop an evaluation method for two-phase flow and heat transfer in a feed-water heater. The following results were obtained by a two-dimensional analysis of a feed-water heater for a power plant. (1) In the model, the film flowed downward in laminar flow due to gravity, with droplet entrainment and deposition. For evaluation of the film thickness, Fujii's equation was used in order to account for forced convection of steam flow. (2) Based on the former experimental data, the droplet deposition coefficient and droplet entrainment rate of liquid film were determined. When the ratio at which the liquid film directly flowed from an upper heat transfer tube to a lower heat transfer tube was 0.7, the calculated total heat transfer rate agreed with the measured value of 130 MW. (3) At the upper region of a heat transfer tube bundle where film thickness was thin, and at the outer region of a heat transfer tube bundle where steam velocity was high, the heat transfer rate was large. (author)

  15. Local and integral ulrasonic gauges for two-phase flow instrumentation in nuclear reactor and safety technology. Final report

    International Nuclear Information System (INIS)

    Chochlov, V.N.; Duncev, A.V.; Ivanov, V.V.; Kontelev, V.V.; Melnikov, V.I.; Stoppel, L.K.; Prasser, H.M.; Zippe, W.; Zschau, J.; Zboray, R.

    2000-11-01

    The present project was executed in the frame of a co-operation agreement between FZR and the scientific group of Prof. Melnikov of the technical University of Nishny Novgorod (TUNN) in the Russian Federation. It is part of the Federal Government's programme for the provision of advice for Eastern Europe on the building up of democracy and social market economy (TRANSFORM Programme). New methods of two-phase flow instrumentation were developed: Intrusive wave-guide probes can be used for local void fraction measurements. The new ultrasonic mesh sensors allow a fast two-phase flow visualisation with about 250 frames per second. Experiments carried out at the test loop in Rossendorf, but also the tests at the DESIRE facility in Delft have shown that both local wave-guide probes and ultrasonic mesh sensors can be successfully applied under the conditions of high pressure and temperature steam-water mixture, as well as in organic liquids and refrigerants. Furthermore, non-intrusive wave-guide sensors as well as density sensors based on the measurement of the wave propagation velocity in wave-guides immersed into the measuring liquid were developed and tested. In the present stage of the development, the non-intrusive sensors can rather be used for a qualitative gas respectively level detection than for void fraction measurements. The wave-guide density sensor was successfully demonstrated that it is able to measure densities of single-phase liquids. It requires further development of the electronic circuitry. The main innovation was achieved by the development of the ultrasonic mesh sensor, the resoluting capability of which is comparable to methods like electrical wire-mesh sensors and ultra-fast X-ray tomography, while the device itself is robust and low expensive. (orig.) [de

  16. Prediction of two-phase choked-flow through safety valves

    International Nuclear Information System (INIS)

    Arnulfo, G; Bertani, C; De Salve, M

    2014-01-01

    Different models of two-phase choked flow through safety valves are applied in order to evaluate their capabilities of prediction in different thermal-hydraulic conditions. Experimental data available in the literature for two-phase fluid and subcooled liquid upstream the safety valve have been compared with the models predictions. Both flashing flows and non-flashing flows of liquid and incondensable gases have been considered. The present paper shows that for flashing flows good predictions are obtained by using the two-phase valve discharge coefficient defined by Lenzing and multiplying it by the critical flow rate in an ideal nozzle evaluated by either Omega Method or the Homogeneous Non-equilibrium Direct Integration. In case of non-flashing flows of water and air, Leung/Darby formulation of the two-phase valve discharge coefficient together with the Omega Method is more suitable to the prediction of flow rate.

  17. Single and two-phase flow pressure drop for CANFLEX bundle

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G R; Bullock, D E [Atomic Energy of Canada Limited, Ontario (Canada)

    1999-12-31

    Friction factor and two-phase flow frictional multiplier for a CANFLEX bundle are newly developed and presented in this paper. CANFLEX as a 43-element fuel bundle has been developed jointly by AECL/KAERI to provide greater operational flexibility for CANDU reactor operators and designers. Friction factor and two-phase flow frictional multiplier have been developed by using the experimental data of pressure drops obtained from two series of Freon-134a (R-134a) CHF tests with a string of simulated CANFLEX bundles in a single phase and a two-phase flow conditions. The friction factor for a CANFLEX bundle is found to be about 20% higher than that of Blasius for a smooth circular pipe. The pressure drop predicted by using the new correlations of friction factor and two-phase frictional multiplier are well agreed with the experimental pressure drop data of CANFLEX bundle within {+-} 5% error. 11 refs., 5 figs. (Author)

  18. Single and two-phase flow pressure drop for CANFLEX bundle

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Ji Su; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E. [Atomic Energy of Canada Limited, Ontario (Canada)

    1998-12-31

    Friction factor and two-phase flow frictional multiplier for a CANFLEX bundle are newly developed and presented in this paper. CANFLEX as a 43-element fuel bundle has been developed jointly by AECL/KAERI to provide greater operational flexibility for CANDU reactor operators and designers. Friction factor and two-phase flow frictional multiplier have been developed by using the experimental data of pressure drops obtained from two series of Freon-134a (R-134a) CHF tests with a string of simulated CANFLEX bundles in a single phase and a two-phase flow conditions. The friction factor for a CANFLEX bundle is found to be about 20% higher than that of Blasius for a smooth circular pipe. The pressure drop predicted by using the new correlations of friction factor and two-phase frictional multiplier are well agreed with the experimental pressure drop data of CANFLEX bundle within {+-} 5% error. 11 refs., 5 figs. (Author)

  19. Two-phase flow experiments in a model of the hot leg of a pressurised water reactor. Technical report

    Energy Technology Data Exchange (ETDEWEB)

    Seidel, Tobias; Vallee, Christophe; Lucas, Dirk; Beyer, Matthias; Deendarlianto

    2011-09-15

    In order to investigate the two-phase flow behaviour in a complex reactor-typical geometry and to supply suitable data for CFD code validation, a model of the hot leg of a pressurised water reactor was built at FZD. The hot leg model is operated in the pressure chamber of the TOPFLOW test facility, which is used to perform high-pressure experiments under pressure equilibrium with the inside atmosphere of the chamber. This technique makes it possible to visualise the two-phase flow through large windows, also at reactor-typical pressure levels. In order to optimise the optical observation possibilities, the test section was designed with a rectangular cross-section. Experiments were performed with air and water at 1.5 and 3.0 bar at room temperature as well as with steam and water at 15, 30 and 50 bar and the corresponding saturation temperature (i.e. up to 264 C). The total of 194 runs are divided into 4 types of experiments covering stationary co-current flow, counter-current flow, flow without water circulation and transient counter-current flow limitation (CCFL) experiments. This report provides a detailed documentation of the experiments including information on the experimental setup, experimental procedure, test matrix and on the calibration of the measuring devices. The available data is described and data sheets were arranged for each experiment in order to give an overview of the most important parameters. For the cocurrent flow experiments, water level histograms were arranged and used to characterise the flow in the hot leg. In fact, the form of the probability distribution was found to be sensitive to the boundary conditions and, therefore, is useful for the CFD comparison. Furthermore, the flooding characteristics of the hot leg model plotted in terms of the classical Wallis parameter or Kutateladze number were found to fail to properly correlate the data of the air/water and steam/water series. Therefore, a modified Wallis parameter is proposed, which

  20. Two-phase flow experiments in a model of the hot leg of a pressurised water reactor. Technical report

    International Nuclear Information System (INIS)

    Seidel, Tobias; Vallee, Christophe; Lucas, Dirk; Beyer, Matthias; Deendarlianto

    2011-09-01

    In order to investigate the two-phase flow behaviour in a complex reactor-typical geometry and to supply suitable data for CFD code validation, a model of the hot leg of a pressurised water reactor was built at FZD. The hot leg model is operated in the pressure chamber of the TOPFLOW test facility, which is used to perform high-pressure experiments under pressure equilibrium with the inside atmosphere of the chamber. This technique makes it possible to visualise the two-phase flow through large windows, also at reactor-typical pressure levels. In order to optimise the optical observation possibilities, the test section was designed with a rectangular cross-section. Experiments were performed with air and water at 1.5 and 3.0 bar at room temperature as well as with steam and water at 15, 30 and 50 bar and the corresponding saturation temperature (i.e. up to 264 C). The total of 194 runs are divided into 4 types of experiments covering stationary co-current flow, counter-current flow, flow without water circulation and transient counter-current flow limitation (CCFL) experiments. This report provides a detailed documentation of the experiments including information on the experimental setup, experimental procedure, test matrix and on the calibration of the measuring devices. The available data is described and data sheets were arranged for each experiment in order to give an overview of the most important parameters. For the cocurrent flow experiments, water level histograms were arranged and used to characterise the flow in the hot leg. In fact, the form of the probability distribution was found to be sensitive to the boundary conditions and, therefore, is useful for the CFD comparison. Furthermore, the flooding characteristics of the hot leg model plotted in terms of the classical Wallis parameter or Kutateladze number were found to fail to properly correlate the data of the air/water and steam/water series. Therefore, a modified Wallis parameter is proposed, which

  1. Numerical study for two phase flow in the near nozzle region of turbine combustors

    International Nuclear Information System (INIS)

    Pervez, K.; Mushtaq, S.

    1999-01-01

    In the present study flow conditions in the near nozzle region of the combustion chamber have been investigated. There exists two-phase flow in this region. The overall performance and pollutant formation in the combustion chamber have been investigated. There exists two-phase flow in this region. The overall performance and pollutant formation in the combustion zone largely depends on the spray field in the near nozzle region the studies are conducted to determined the effects of multi jets on the flow pattern in the near nozzle region The phase doppler particle analyzer (PDPA) has been used to measure the velocities and sizes of the droplets. The flow field of two-phase liquid drop-air jets is formed from three injectors arranged in t line. Furthermore the two-phase flow field has been analyzed numerically also. The numerical analysis consists of two computational models, namely (i) 3 non-evaporating two-phase jets, (II) 3 evaporating two phase jets. The Eulerian-Eulerian approach in incorporated in both the numerical models. Since the flow is turbulent, a two-equation model (k-Epsilon) is implemented in the numerical analysis. Numerical solution of the conservation equation is obtained using PHOENICS computer code. Boundary conditions are provided from the experimental measurements. Numerical domain for the two models of the analysis starts at some distance (about 10 diameters of the injector orifice) where the atomization process is complete and droplet size and velocity could be measured experimentally. (author)

  2. Theory and design of heat exchanger : Double pipe and heat exchanger in abnormal condition

    International Nuclear Information System (INIS)

    Min, Ui Dong

    1996-02-01

    This book introduces theory and design of heat exchanger, which includes HTRI program, multiple tube heat exchanger external heating, theory of heat transfer, basis of design of heat exchanger, two-phase flow, condensation, boiling, material of heat exchanger, double pipe heat exchanger like hand calculation, heat exchanger in abnormal condition such as Jackets Vessel, and Coiled Vessel, design and summary of steam tracing.

  3. Two-phase-flow cooling concept for fusion reactor blankets

    International Nuclear Information System (INIS)

    Bender, D.J.; Hoffman, M.A.

    1977-01-01

    The new two-phase heat transfer medium proposed is a mixture of potassium droplets and helium which permits blanket operation at hih temperature and low pressure, while maintaining acceptable pumping power requirements, coolant ducting size, and blanket structure fractions. A two-phase flow model is described. The helium pumping power and the primary heat transfer loop are discussed

  4. Unsteady State Two Phase Flow Pressure Drop Calculations

    OpenAIRE

    Ayatollahi, Shahaboddin

    1992-01-01

    A method is presented to calculate unsteady state two phase flow in a gas-liquid line based on a quasi-steady state approach. A computer program for numerical solution of this method was prepared. Results of calculations using the computer program are presented for several unsteady state two phase flow systems

  5. Steam generators

    International Nuclear Information System (INIS)

    Hayden, R.L.J.

    1979-01-01

    Steam generators for nuclear reactors are designed so that deposition of solids on the surface of the inlet side of the tubesheet or the inlet header with the consequent danger of corrosion and eventual tube failure is obviated or substantially reduced. (U.K.)

  6. A Complex Solar Coronal Jet with Two Phases

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Jie; Su, Jiangtao; Deng, Yuanyong [Key Laboratory of Solar Activity, National Astronomical Observatories, Chinese Academy of Sciences, Beijing 100012 (China); Priest, E. R., E-mail: chenjie@bao.ac.cn [Mathematical Institute, University of St Andrews, North Haugh, St Andrews, KY16 9SS (United Kingdom)

    2017-05-01

    Jets often occur repeatedly from almost the same location. In this paper, a complex solar jet was observed with two phases to the west of NOAA AR 11513 on 2012 July 2. If it had been observed at only moderate resolution, the two phases and their points of origin would have been regarded as identical. However, at high resolution we find that the two phases merge into one another and the accompanying footpoint brightenings occur at different locations. The phases originate from different magnetic patches rather than being one phase originating from the same patch. Photospheric line of sight (LOS) magnetograms show that the bases of the two phases lie in two different patches of magnetic flux that decrease in size during the occurrence of the two phases. Based on these observations, we suggest that the driving mechanism of the two successive phases is magnetic cancellation of two separate magnetic fragments with an opposite-polarity fragment between them.

  7. Dynamics of the free surface of stratified two-phase flows in channels with rectangular cross-sections

    International Nuclear Information System (INIS)

    Vallee, Christophe

    2012-01-01

    Stratified two-phase flows were investigated at different test facilities with horizontal test sections in order to provide an experimental database for the development and validation of computational fluid dynamics (CFD) codes. These channels were designed with rectangular cross-sections to enable optimal observation conditions for the application of optical measurement techniques. Consequently, the local flow structure was visualised with a high-speed video camera, delivering data with highresolution in space and time as needed for CFD code validation. Generic investigations were performed at atmospheric pressure and room temperature in two air/water channels made of acrylic glass. Divers preliminary experiments were conducted with various measuring systems in a test section mounted between two separators. The second test facility, the Horizontal Air/Water Channel (HAWAC), is dedicated to co-current flow investigations. The hydraulic jump as the quasi-stationary discontinuous transition between super- and subcritical flow was studied in this closed channel. Moreover, the instable wave growth leading to slug flow was investigated from the test section inlet. For quantitative analysis of the optical measurements, an algorithm was developed to recognise the stratified interface in the camera frames, allowing statistical treatments for comparison with CFD calculation results. The third test apparatus was installed in the pressure chamber of the TOPFLOW test facility in order to be operated at reactor typical conditions under pressure equilibrium with the vessel atmosphere. The test section representing a flat model of the hot leg of the German Konvoi pressurised water reactor (PWR) scaled at 1:3 is equipped with large glass side walls in the region of the elbow and of the steam generator inlet chamber to allow visual observations. The experiments were conducted with air and water at room temperature and maximum pressures of 3 bar as well as with steam and water at

  8. Dynamics of the free surface of stratified two-phase flows in channels with rectangular cross-sections

    Energy Technology Data Exchange (ETDEWEB)

    Vallee, Christophe

    2012-08-22

    Stratified two-phase flows were investigated at different test facilities with horizontal test sections in order to provide an experimental database for the development and validation of computational fluid dynamics (CFD) codes. These channels were designed with rectangular cross-sections to enable optimal observation conditions for the application of optical measurement techniques. Consequently, the local flow structure was visualised with a high-speed video camera, delivering data with highresolution in space and time as needed for CFD code validation. Generic investigations were performed at atmospheric pressure and room temperature in two air/water channels made of acrylic glass. Divers preliminary experiments were conducted with various measuring systems in a test section mounted between two separators. The second test facility, the Horizontal Air/Water Channel (HAWAC), is dedicated to co-current flow investigations. The hydraulic jump as the quasi-stationary discontinuous transition between super- and subcritical flow was studied in this closed channel. Moreover, the instable wave growth leading to slug flow was investigated from the test section inlet. For quantitative analysis of the optical measurements, an algorithm was developed to recognise the stratified interface in the camera frames, allowing statistical treatments for comparison with CFD calculation results. The third test apparatus was installed in the pressure chamber of the TOPFLOW test facility in order to be operated at reactor typical conditions under pressure equilibrium with the vessel atmosphere. The test section representing a flat model of the hot leg of the German Konvoi pressurised water reactor (PWR) scaled at 1:3 is equipped with large glass side walls in the region of the elbow and of the steam generator inlet chamber to allow visual observations. The experiments were conducted with air and water at room temperature and maximum pressures of 3 bar as well as with steam and water at

  9. Two-phase Flow Ejector as Water Refrigerant by Using Waste Heat

    International Nuclear Information System (INIS)

    Yamanaka, H; Nakagawa, M

    2013-01-01

    Energy saving and the use of clean energy sources have recently become significant issues. It is expected that clean energy sources such as solar panels and fuel cells will be installed in many private dwellings. However, when electrical power is generated, exhaust heat is simultaneously produced. Especially for the summer season, the development of refrigeration systems that can use this waste heat is highly desirable. One approach is an ejector that can reduce the mechanical compression work required in a normal refrigeration cycle. We focus on the use of water as a refrigerant, since this can be safely implemented in private dwellings. Although the energy conversion efficiency is low, it is promising because it can use heat that would otherwise be discarded. However, a steam ejector refrigeration cycle requires a large amount of energy to change saturated water into vapour. Thus, we propose a more efficient two-phase flow ejector cycle. Experiments were carried out in which the quality of the two-phase flow from a tank was varied, and the efficiency of the ejector and nozzle was determined. The results show that a vacuum state can be achieved and suction exerted with a two-phase flow state at the ejector nozzle inlet.

  10. Numerical methods for limit problems in two-phase flow models

    International Nuclear Information System (INIS)

    Cordier, F.

    2011-01-01

    Numerical difficulties are encountered during the simulation of two-phase flows. Two issues are studied in this thesis: the simulation of phase transitions on one hand, and the simulation of both compressible and incompressible flows in the other hand. Un asymptotic study has shown that the loss of hyperbolicity of the bi fluid model was responsible for the difficulties encountered by the Roe scheme during the simulation of phase transitions. Robust and accurate polynomial schemes have thus been developed. To tackle the occasional lack of positivity of the solution, a numerical treatment based on adaptive diffusion was proposed and allowed to simulate with accuracy the test-cases of a boiling channel with creation of vapor and a tee-junction with separation of the phases. In a second part, an all-speed scheme for compressible and incompressible flows have been proposed. This pressure-based semi-implicit asymptotic preserving scheme is conservative, solves an elliptic equation on the pressure, and has been designed for general equations of state. The scheme was first developed for the full Euler equations and then extended to the Navier-Stokes equations. The good behaviour of the scheme in both compressible and incompressible regimes have been investigated. An extension of the scheme to the two-phase mixture model was implemented and demonstrated the ability of the scheme to simulate two-phase flows with phase change and a water-steam equation of state. (author) [fr

  11. Investigation of two-phase liquid-metal magnetohydrodynamic power systems

    International Nuclear Information System (INIS)

    Amend, W.E.; Fabris, G.; Cutting, J.

    1975-01-01

    A two-phase Liquid-Metal MHD (LMMHD) system is under development at the Argonne National Laboratory, and results are presented for detailed cycle analysis and systems studies, the experimental facility, and the thermal and magneto fluid mechanics problems encountered. The studies indicate that the LMMHD cycle will operate efficiently in the temperature range of 1000-1600 0 F (50 percent efficiency with a maximum cycle temperature of 1600 0 F) and is therefore potentially compatible with many advanced heat sources under development such as the LMFBR, fluidized-bed coal combustor, HTGCR and the fusion reactor. Of special interest is the coupling to the LMFBR thereby eliminating the costly, potentially hazardous liquid-metal/water interface. The results of detailed parametric studies of the heat transfer interfaces between an LMMHD power cycle and an LMFBR and a steam bottoming plant are described. Experimental evaluation of the two-phase LMMHD generator was performed in an ambient temperature NaK--N 2 facility at ANL. Results of these experiments, performed to determine the operating characteristics of the device as a function of the various independent parameters and to investigate two-phase flow, are given. (U.S.)

  12. Two-phase flow heat transfer in nuclear reactor systems

    International Nuclear Information System (INIS)

    Koncar, Bostjan; Krepper, Eckhard; Bestion, Dominique; Song, Chul-Hwa; Hassan, Yassin A.

    2013-01-01

    occurring in at least two different spatial scales. Uncertainty in modelling of bubble departure diameter at boiling was studied by M. Matkovic and B. Koncar. In this article the propagation of input uncertainties for the simplified model of bubble departure size is evaluated. A methodology for estimating the prediction capability of a given correlation is provided taking into account its range of applicability. Aqueous nanofluids have a great potential for cooling applications, hence they have been studied in the article of P.N. Alekseev et al. as a possible coolant in pressurized water reactor (PWR). The theoretical study presents how a stable formation of nanoparticles in water solution can be established. Formation of fractal nanoparticles with a higher thermal conductivity than water can enhance the heat transfer of water used as a coolant in PWR. Apart from solid particles, also alternative formation of gaseous nanoparticles in density fluctuations of water is discussed. The article of R. Rzehak and E. Krepper provides a comprehensive overview of the state-of-the-art in the field of CFD modelling of subcooled flow boiling. The efficient predictive capability of current models requires calibration of model parameters over a wide range of measured data and operating conditions.The results presented in the article confirmed the great potential of the existing modelling approach for the 3D simulation of subcooled flow boiling in industrial applications but also highlight the need for specific model improvements to achieve highly accurate predictions. Two articles deal with one-dimensional analyses of two phase flows. In the article of O.Costa et al., a rapid depressurization in vertical heated pipe is simulated with the in-house 1D computer code WAHA, which was developed specifically for simulations of two-phase water hammer phenomena. The WAHA results were confronted with the simulations of the well-known system code RELAP5 on the same experimental data. The thermal

  13. Investigation of a two-phase nozzle flow and validation of several computer codes by the experimental data

    International Nuclear Information System (INIS)

    Kedziur, F.

    1980-03-01

    Stationary experiments with a convergent nozzle are performed in order to validate advanced two-phase computer codes, which find application in the blowdown-phase of a loss-of-coolant accident (LOCA). The steam/water flow presents a broad variety of initial conditions: The pressure varies between 2 and 13 MPa, the void fraction between 0 (subcooled) and about 80%, a great number of subcritical as well as critical experiments with different flow pattern is investigated. Additional air/water experiments serve for the separation of phase transition effects. The transient acceleration of the fluid in the LOCA-case is simulated by a local acceleration in the experiments. The layout of the nozzle and the applied measurement technique allow for a separate testing of physical models and the determination of empirical model parameters, respectively: In the four codes DUESE, DRIX-20, RELAP4/MOD6 and STRUYA the models - if they exist - for slip between the phases, thermodynamic non-equilibrium, pipe friction and critical mass flow rate are validated and criticised in comparison with the experimental data, and the corresponding model parameters are determined. The parameters essentially are a function of the void fraction. (orig.) [de

  14. Model and control scheme for recirculation mode direct steam generation parabolic trough solar power plants

    International Nuclear Information System (INIS)

    Guo, Su; Liu, Deyou; Chen, Xingying; Chu, Yinghao; Xu, Chang; Liu, Qunming; Zhou, Ling

    2017-01-01

    Highlights: •A nonlinear dynamic model of recirculation DSG parabolic trough is developed. •Collector row, water separator and spray attemperator are modeled, respectively. •The dynamic behaviors of the collector field are simulated and analyzed. •Transfer functions of water level and outlet fluid temperature are derived. •Multi-model switching generalized predictive control strategy is developed. -- Abstract: This work describes and evaluates a new nonlinear dynamic model, and a new generalized predictive control scheme for a collector field of direct steam generation parabolic troughs in recirculation mode. Modeling the dynamic behaviors of collector fields is essential to design, testing and validation of automatic control systems for direct steam generation parabolic troughs. However, the behaviors of two-phase heat transfer fluids impose challenges to simulating and developing process control schemes. In this work, a new nonlinear dynamic model is proposed, based on the nonlinear distributed parameter and the nonlinear lumped parameter methods. The proposed model is used to simulate and analyze the dynamic behaviors of the entire collector field for recirculation mode direct steam generation parabolic troughs under different weather conditions, without excessive computational costs. Based on the proposed model, transfer functions for both the water level of the separator and outlet steam temperatures are derived, and a new multi-model switching generalized predictive control scheme is developed for simulated control of the plant behaviors for a wide region of operational conditions. The proposed control scheme achieves excellent control performance and robustness for systems with long delay, large inertia and time-varying parameters, and efficiently solves the model mismatching problem in direct steam generation parabolic troughs. The performances of the model and control scheme are validated with design data from the project of Integration of Direct

  15. Industrial aspects of gas-liquid two-phase flow

    International Nuclear Information System (INIS)

    Hewitt, G.F.

    1977-01-01

    The lecture begins by reviewing the various types of plant in which two phase flow occurs. Specifically, boiling plant, condensing plant and pipelines are reviewed, and the various two phase flow problems occurring in them are described. Of course, many other kinds of chemical engineering plant involve two phase flow, but are somewhat outside the scope of this lecture. This would include distillation columns, vapor-liquid separators, absorption towers etc. Other areas of industrial two phase flow which have been omitted for space reasons from this lecture are those concerned with gas/solids, liquid/solid and liquid/liquid flows. There then follows a description of some of the two phase flow processes which are relevant in industrial equipment and where special problems occur. The topics chosen are as follows: (1) pressure drop; (2) horizontal tubes - separation effects non-uniformites in heat transfer coefficient, effect of bends on dryout; (3) multicomponent mixtures - effects in pool boiling, mass transfer effects in condensation and Marangoni effects; (4) flow distribution - manifold problems in single phase flow, separation effects at a single T-junction in two phase flow and distribution in manifolds in two phase flow; (5) instability - oscillatory instability, special forms of instability in cryogenic systems; (6) nucleate boiling - effect of variability of surface, unresolved problems in forced convective nucleate boiling; and (7) shell side flows - flow patterns, cross flow boiling, condensation in cross flow

  16. Modeling two-phase flow in PEM fuel cell channels

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yun; Basu, Suman; Wang, Chao-Yang [Electrochemical Engine Center (ECEC), and Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States)

    2008-05-01

    This paper is concerned with the simultaneous flow of liquid water and gaseous reactants in mini-channels of a proton exchange membrane (PEM) fuel cell. Envisaging the mini-channels as structured and ordered porous media, we develop a continuum model of two-phase channel flow based on two-phase Darcy's law and the M{sup 2} formalism, which allow estimate of the parameters key to fuel cell operation such as overall pressure drop and liquid saturation profiles along the axial flow direction. Analytical solutions of liquid water saturation and species concentrations along the channel are derived to explore the dependences of these physical variables vital to cell performance on operating parameters such as flow stoichiometric ratio and relative humility. The two-phase channel model is further implemented for three-dimensional numerical simulations of two-phase, multi-component transport in a single fuel-cell channel. Three issues critical to optimizing channel design and mitigating channel flooding in PEM fuel cells are fully discussed: liquid water buildup towards the fuel cell outlet, saturation spike in the vicinity of flow cross-sectional heterogeneity, and two-phase pressure drop. Both the two-phase model and analytical solutions presented in this paper may be applicable to more general two-phase flow phenomena through mini- and micro-channels. (author)

  17. Identification of relevant conditions and experiments for fuel-coolant interactions in nuclear power plants - SERENA Co-ordinated Programme (Steam Explosion Resolution for Nuclear Applications) Phase 1, Task 1 - Final report

    International Nuclear Information System (INIS)

    Magallon, D.; Scott de Martinville, E.; Chaumont, B.; Filippi, M.; Meignen, R.; Berthoud, G.; Ratel, G.; Melikhov, O.I.; Melikhov, V.I.; Jacobs, H.; Buerger, Manfred; Buck, Michael; Moriyama, K.; Nakamura, H.; Hirano, M.; Muramatsu, K.; Ishikawa, J.; Song, Jinho; Bang, Kwanghyun; Suh, Namduk; Sairanen, Risto; Lindholm, Ilona

    2004-01-01

    SERENA (Steam Explosion Resolution for Nuclear Applications) is an international OECD programme for the resolution of FCI remaining issues. The programme has origin the concerns expressed by the Senior Group Experts on Nuclear Safety Research and Programme (SESAR/FAP) about de-emphasis of FCI research all over the world, while uncertainties still exist on some aspects of FCI. After an evaluation of remaining needs by FCI experts in a meeting October 2000, a proposal was matured during 2001 following CSNI recommendations that existing knowledge should be carefully assessed before carrying out new experiments, and reactor application should be the focus of any new action. The work programme was approved by CSNI December 2001. The programme started January 2002. The overall objective of SERENA is to obtain convergence on the understanding of FCI processes and energetics, as well as on method(s) for reliable estimate of the magnitude of loadings for realistic reactor conditions, in order to bring understanding and predictability of FCI energetics to desirable levels for risk management. The work is performed in two phases: - Phase 1 analyses and evaluates knowledge and data on FCI by using available tools with the aim to identify areas where large uncertainties/ discrepancies still subsist and are important for predicting loads in reactors with a sufficient level of confidence, and work to be done, if any, to reduce these uncertainties/discrepancies. - Phase 2 will implement analytical and experimental actions to resolve these uncertainties/ discrepancies, if required. The objective of Phase 1 is reached through comparative calculations by available tools of existing experiments and reactor cases. It is divided into five tasks: 1. Identification of relevant conditions for FCI in reactor, 2. Comparison of various approaches for calculating jet break-up and pre-mixing, 3. Comparison of various approaches for calculating the explosion phase, 4. Reactor applications, 5

  18. Regimes of Two-Phase Flow in Short Rectangular Channel

    Science.gov (United States)

    Chinnov, Evgeny A.; Guzanov, Vladimir V.; Cheverda, Vyacheslav; Markovich, Dmitry M.; Kabov, Oleg A.

    2009-08-01

    Experimental study of two-phase flow in the short rectangular horizontal channel with height 440 μm has been performed. Characteristics of liquid motion inside the channel have been registered and measured by the Laser Induced Fluorescence technique. New information has allowed determining more precisely the characteristics of churn regime and boundaries between different regimes of two-phase flow. It was shown that formation of some two-phase flow regimes and transitions between them are determined by instability of the flow in the lateral parts of the channel.

  19. What types of investors generate the two-phase phenomenon?

    Science.gov (United States)

    Ryu, Doojin

    2013-12-01

    We examine the two-phase phenomenon described by Plerou, Gopikrishnan, and Stanley (2003) [1] in the KOSPI 200 options market, one of the most liquid options markets in the world. By analysing a unique intraday dataset that contains information about investor type for each trade and quote, we find that the two-phase phenomenon is generated primarily by domestic individual investors, who are generally considered to be uninformed and noisy traders. In contrast, our empirical results indicate that trades by foreign institutions, who are generally considered informed and sophisticated investors, do not exhibit two-phase behaviour.

  20. Thermo-Fluid Dynamics of Two-Phase Flow

    CERN Document Server

    Ishii, Mamrou

    2011-01-01

    "Thermo-fluid Dynamics of Two-Phase Flow, Second Edition" is focused on the fundamental physics of two-phase flow. The authors present the detailed theoretical foundation of multi-phase flow thermo-fluid dynamics as they apply to: Nuclear reactor transient and accident analysis; Energy systems; Power generation systems; Chemical reactors and process systems; Space propulsion; Transport processes. This edition features updates on two-phase flow formulation and constitutive equations and CFD simulation codes such as FLUENT and CFX, new coverage of the lift force model, which is of part

  1. Three-Dimensional Modeling of a Steam-Line Break in a Boiling Water Reactor

    International Nuclear Information System (INIS)

    Tinoco, Hernan

    2002-01-01

    Because of weld problems, the core grids of Units 1 and 2 at the Forsmark nuclear power plant have been replaced by grids of a new design, consisting of a single machined piece without welds. The qualifying structural analysis has been carried out considering dynamic loads, which implies that even loss-of-coolant accidents have to be included. Therefore, a detailed time description of the loads acting on the different internal parts of the reactor is needed. To achieve sufficient space and time resolution, a computational fluid dynamics (CFD) analysis was considered to be a viable alternative.A CFD analysis of a steam-line break in the boiling water reactor of Unit 2 is the subject of this work. The study is based on the assumption that the timescale of the transient analysis is smaller than the relaxation time of the water-steam system. Therefore, a simulation of only the upper, steam part of the reactor with no two-phase effects (flashing) is feasible.The results obtained display a rather complex behavior of the decompression process, forcing the analysis of the pressure field to be accomplished through animation. In contrast, the computed instantaneous forces over different internal parts oscillate regularly and are approximately twice the forces estimated in the past by simpler methods, with frequencies of 30 to 40 Hz; top amplitudes of ∼1.64 MN; and relatively low damping, ∼25% after 0.5 s.According to the present results, this type of modeling is physically meaningful for simulation timescales smaller than the water-steam relaxation time, i.e., ∼0.5 s at reactor conditions. At larger times, a two-phase model is necessary to describe the decompression process since two-phase effects are dominant. The results have not yet been validated with experiments, but validation computations will be run in the future for comparison with results of the Marviken tests

  2. Analysis of Two Phase Natural Circulation Flow in the Cooling Channel of the PECS

    Energy Technology Data Exchange (ETDEWEB)

    Park, R. J; Ha, K. S; Rhee, B. W; Kim, H. Y [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Decay heat and sensible heat of the relocated and spread corium are removed by the natural circulation flow at the bottom and side wall of the core catcher and the top water cooling of the corium. The coolant in the inclined channel absorbs the decay heat and sensible heat transferred from the corium through the structure of the core catcher body and flows up to the pool as a two phase mixture. On the other hand, some of the pool water will flow into the inlet of the downcomer piping, and will flow into the inclined cooling channel of the core catcher by gravity. As shown in Fig. 1, the engineered cooling channel is designed to provide effective long-term cooling and stabilization of the corium mixture in the core catcher body while facilitating steam venting in the PECS. To maintain the integrity of the ex-vessel core catcher, however, it is necessary that the coolant be sufficiently circulated along the inclined cooling channel to avoid CHF (Critical Heat Flux) on the heating surface of the cooling channel. For this reason, a verification experiment on the cooling capability of the EU-APR1400 core catcher has been performed in the CE (Cooling Experiment)-PECS facility at KAERI. Preliminary simulations of two-phase natural circulation in the CE-PECS were performed to predict two-phase flow characteristics and to determine the natural circulation mass flow rate in the flow channel. In this study, simulations of two-phase natural circulation in a real core catcher of the PECS have been performed to determine the natural circulation mass flow rate in the flow channel using the RELAP5/MOD3 computer code.

  3. Design and development of drag-disc flow meter for measurement of transient two-phase flow

    International Nuclear Information System (INIS)

    Sreenivas Rao, G.; Kukreja, V.; Dolas, P.K.; Venkat Raj, V.

    1989-01-01

    Experiments have been carried out to test the suitability of drag-disc flowmeter for measuring two-phase flow. Calibration tests carried out under single-phase and two-phase flow conditions have confirmed the suitability of the drag-disc flowmeter. The experimental work and the results obtained are presented and discussed in the paper. (author). 6 figs

  4. Two-phase systems. Fundamentals and industrial applications

    International Nuclear Information System (INIS)

    Woillez, Jacques

    2014-01-01

    Two-phase flows are omnipresent in industrial processes in different sectors with the behaviour and control of non-mixing mixtures of gas and liquids, of several liquids, of solids and fluids which are present in the production of raw materials, in the environment, in energy production, in chemistry, in pharmaceutical or food industry. The author presents the fundamentals elements which are needed to perform hardware predictive calculations and to understand typical phenomena associated with these flows. The chapters address fluids mechanics (movement equations, Bernoulli equation, load losses, turbulence, heat exchange coefficients, thermodynamics, compressible flows), two-phase systems (characteristic values, modes of appearance of two-phase flows, conduct flows, suspension mechanics, mass transfers, similarity, numerical simulation), the applications (energy production, agitation and mixing, phase separation, sprays), and peculiar phenomena (Marangoni effect, the tea cup effect, entry jets, water hammer effect, sound speed, two-phase pumping, fluidization)

  5. Visual Analysis of Inclusion Dynamics in Two-Phase Flow.

    Science.gov (United States)

    Karch, Grzegorz Karol; Beck, Fabian; Ertl, Moritz; Meister, Christian; Schulte, Kathrin; Weigand, Bernhard; Ertl, Thomas; Sadlo, Filip

    2018-05-01

    In single-phase flow visualization, research focuses on the analysis of vector field properties. In two-phase flow, in contrast, analysis of the phase components is typically of major interest. So far, visualization research of two-phase flow concentrated on proper interface reconstruction and the analysis thereof. In this paper, we present a novel visualization technique that enables the investigation of complex two-phase flow phenomena with respect to the physics of breakup and coalescence of inclusions. On the one hand, we adapt dimensionless quantities for a localized analysis of phase instability and breakup, and provide detailed inspection of breakup dynamics with emphasis on oscillation and its interplay with rotational motion. On the other hand, we present a parametric tightly linked space-time visualization approach for an effective interactive representation of the overall dynamics. We demonstrate the utility of our approach using several two-phase CFD datasets.

  6. Metrology of two-phase flow: different methods

    International Nuclear Information System (INIS)

    Delhaye, J.M.; Galaup, J.P.; Reocreux, M.; Ricque, R.

    Nine papers are presented concerning different methods of measuring two-phase flow. Some of the methods and equipment discussed include: radiation absorption, electromagnetic flowmeter, anemometry, resistance probes, phase indicating microthermocouples, optical probes, sampling methods, and pitot tubes

  7. Qualitative behaviour of incompressible two-phase flows with phase ...

    Indian Academy of Sciences (India)

    Jan Prüss

    2017-11-07

    Nov 7, 2017 ... Qualitative behaviour of incompressible two-phase flows with phase ... Germany. 2Graduate School of Human and Environmental Studies, Kyoto University, ... Note that j is a dummy variable as it can be eliminated from the ...

  8. Biogasification of solid wastes by two-phase anaerobic fermentation

    International Nuclear Information System (INIS)

    Ghosh, S.; Vieitez, E.R.; Liu, T.; Kato, Y.

    1997-01-01

    Municipal, industrial and agricultural solid wastes, and biomass deposits, cause large-scale pollution of land and water. Gaseous products of waste decomposition pollute the air and contribute to global warming. This paper describes the development of a two-phase fermentation system that alleviates methanogenic inhibition encountered with high-solids feed, accelerates methane fermentation of the solid bed, and captures methane (renewable energy) for captive use to reduce global warming. The innovative system consisted of a solid bed reactor packed with simulated solid waste at a density of 160 kg/m 3 and operated with recirculation of the percolated culture (bioleachate) through the bed. A rapid onset of solids hydrolysis, acidification, denitrification and hydrogen gas formation was observed under these operating conditions. However, these fermentative reactions stopped at a total fatty acids concentration of 13,000 mg/l (as acetic) at pH 5, with a reactor head-gas composition of 75 percent carbon dioxide, 20 percent nitrogen, 2 percent hydrogen and 3 percent methane. Fermentation inhibition was alleviated by moving the bioleachate to a separate methane-phase fermenter, and recycling methanogenic effluents at pH 7 to the solid bed. Coupled operation of the two reactors promoted methanogenic conversion of the high-solids feed. (author)

  9. Passive Two-Phase Cooling of Automotive Power Electronics: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Moreno, G.; Jeffers, J. R.; Narumanchi, S.; Bennion, K.

    2014-08-01

    Experiments were conducted to evaluate the use of a passive two-phase cooling strategy as a means of cooling automotive power electronics. The proposed cooling approach utilizes an indirect cooling configuration to alleviate some reliability concerns and to allow the use of conventional power modules. An inverter-scale proof-of-concept cooling system was fabricated, and tests were conducted using the refrigerants hydrofluoroolefin HFO-1234yf and hydrofluorocarbon HFC-245fa. Results demonstrated that the system can dissipate at least 3.5 kW of heat with 250 cm3 of HFC-245fa. An advanced evaporator design that incorporates features to improve performance and reduce size was conceived. Simulation results indicate its thermal resistance can be 37% to 48% lower than automotive dual side cooled power modules. Tests were also conducted to measure the thermal performance of two air-cooled condensers--plain and rifled finned tube designs. The results combined with some analysis were then used to estimate the required condenser size per operating conditions and maximum allowable system (i.e., vapor and liquid) temperatures.

  10. Dielectric barrier discharge in a two-phase mixture

    Energy Technology Data Exchange (ETDEWEB)

    Ye Qizheng; Zhang Ting; Lu Fei; Li Jin; He Zhenghao; Lin Fuchang [College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2008-01-21

    This paper reports the experimental investigation of the dielectric barrier discharge in which the gap area is filled with a two-phase mixture (TPM), air and solid particles. We found that there are two kinds of discharges in the TPM. One is the surface discharge generated on the surface of the solid particles and the other is the filament discharge generated in the air void. For the case of low volume fraction of solid particles, the surface discharge starts to occur when the applied voltage is higher than the onset voltage. At a further voltage increase, the filament discharge takes place at the same time. For the case of high volume fraction, such as the packed-bed reactor, only the surface discharge exists. Under the condition of the same volume fraction, the larger the diameter of the solid particles, the lower the surface discharge onset voltage. As a conclusion, we think that the plasma reactor using the form of low volume fraction of solid particles may be a better choice for waste-gas treatment enhanced by catalysts.

  11. Abnormal breakdown characteristic in a two-phase mixture

    International Nuclear Information System (INIS)

    Ye Qizheng; Li Jin; Lu Fei

    2006-01-01

    A two-phase mixture (TPM) is a mixture of gas and macroparticles of high concentration. Based on Townsend's theory, a new cell-iterative model in analytical form for the breakdown mechanism in TPM is presented. Compared with the original cell-iterative model in our previous paper, the obstructive factor of the macroparticles that influences the electron avalanche propagation is considered, except for the macroparticles distorting the electrical field and capture of the electrons. The cell attractive parameter k is presented according to the classical continuum theory for field charging. The modified Paschen law for a TPM is presented to calculate the breakdown voltage. The breakdown voltage of the TPM, U TPM , increases gradually with an increase in the macroparticle number density (m). The voltage U TPM is lower than that of the pure gas at low m values and larger at high m values. With a decrease of the macroparticle volume fraction and the dielectric mismatch, the voltage U TPM increases gradually at low m values and decreases gradually at high m values. The voltage U TPM at pd 200 cm Torr is lower than that at pd = 760 cm Torr for low m values and larger for high m values. This kind of abnormal breakdown characteristic in the TPM occurs in the case of high macroparticle volume fraction. On the other hand, the minimum of the TPM's Paschen curve increases with increase in m. It provides the possibility and the conditions of greatly increasing the breakdown voltage in a nearly uniform field

  12. Two-phase flow patterns in horizontal rectangular minichannel

    Directory of Open Access Journals (Sweden)

    Ron’shin Fedor

    2016-01-01

    Full Text Available The two-phase flow in a short horizontal channel of rectangular cross-section of 1 × 19 mm2 has been studied experimentally. Five conventional two-phase flow patterns have been detected (bubble, churn, stratified, annular and jet and transitions between them have been determined. It is shown that a change in the width of the horizontal channels has a substantial effect on the boundaries between the flow regimes.

  13. Refrigeration. Two-Phase Flow. Flow Regimes and Pressure Drop

    DEFF Research Database (Denmark)

    Knudsen, Hans-Jørgen Høgaard

    2002-01-01

    The note gives the basic definitions used in two-phase flow. Flow regimes and flow regimes map are introduced. The different contributions to the pressure drop are stated together with an imperical correlation from the litterature.......The note gives the basic definitions used in two-phase flow. Flow regimes and flow regimes map are introduced. The different contributions to the pressure drop are stated together with an imperical correlation from the litterature....

  14. Two-phase cooling fluids; Les fluides frigoporteurs diphasiques

    Energy Technology Data Exchange (ETDEWEB)

    Lallemand, A. [Institut National des Sciences Appliquees (INSA), 69 - Lyon (France)

    1997-12-31

    In the framework of the diminution of heat transfer fluid consumption, the concept of indirect refrigerating circuits, using cooling intermediate fluids, is reviewed and the fluids that are currently used in these systems are described. Two-phase cooling fluids advantages over single-phase fluids are presented with their thermophysical characteristics: solid fraction, two-phase mixture enthalpy, thermal and rheological properties, determination of heat and mass transfer characteristics, and cold storage through ice slurry

  15. Review on two-phase flow instabilities in narrow spaces

    International Nuclear Information System (INIS)

    Tadrist, L.

    2007-01-01

    Instabilities in two-phase flow have been studied since the 1950s. These phenomena may appear in power generation and heat transfer systems where two-phase flow is involved. Because of thermal management in small size systems, micro-fluidics plays an important role. Typical processes must be considered when the channel hydraulic diameter becomes very small. In this paper, a brief review of two-phase flow instabilities encountered in channels having hydraulic diameters greater than 10 mm are presented. The main instability types are discussed according to the existing experimental results and models. The second part of the paper examines two-phase flow instabilities in narrow spaces. Pool and flow boiling cases are considered. Experiments as well as theoretical models existing in the literature are examined. It was found that several experimental works evidenced these instabilities meanwhile only limited theoretical developments exist in the literature. In the last part of the paper an interpretation of the two-phase flow instabilities linked to narrow spaces are presented. This approach is based on characteristic time scales of the two-phase flow and bubble growth in the capillaries

  16. Steam turbine cycle

    International Nuclear Information System (INIS)

    Okuzumi, Naoaki.

    1994-01-01

    In a steam turbine cycle, steams exhausted from the turbine are extracted, and they are connected to a steam sucking pipe of a steam injector, and a discharge pipe of the steam injector is connected to an inlet of a water turbine. High pressure discharge water is obtained from low pressure steams by utilizing a pressurizing performance of the steam injector and the water turbine is rotated by the high pressure water to generate electric power. This recover and reutilize discharged heat of the steam turbine effectively, thereby enabling to improve heat efficiency of the steam turbine cycle. (T.M.)

  17. Entropy analysis on non-equilibrium two-phase flow models

    International Nuclear Information System (INIS)

    Karwat, H.; Ruan, Y.Q.

    1995-01-01

    A method of entropy analysis according to the second law of thermodynamics is proposed for the assessment of a class of practical non-equilibrium two-phase flow models. Entropy conditions are derived directly from a local instantaneous formulation for an arbitrary control volume of a structural two-phase fluid, which are finally expressed in terms of the averaged thermodynamic independent variables and their time derivatives as well as the boundary conditions for the volume. On the basis of a widely used thermal-hydraulic system code it is demonstrated with practical examples that entropy production rates in control volumes can be numerically quantified by using the data from the output data files. Entropy analysis using the proposed method is useful in identifying some potential problems in two-phase flow models and predictions as well as in studying the effects of some free parameters in closure relationships

  18. Numerical analysis of critical two-phase flow in a convergent-divergent nozzle

    International Nuclear Information System (INIS)

    Romstedt, P.; Werner, W.

    1985-01-01

    The numerical calculation of critical two-phase flow in a convergent-divergent nozzle is complicated by a singularity of the fluid flow equations at the unknown critical point. This paper describes a method which is able to calculate critical state and its location without any additional assumptions. The critical state is identified by its mathematical properties: characteristics and solvability of linear systems with singular matrix. Because the numerically evaluable mathematical properties are only necessary conditions for the existence of critical flow, some physical ''compatibility-criteria'' (flow velocity equals two-phase sonic velocity, critical flow is independent of downstream flow state variations) are used as a substitute for mathematically sufficient conditions. Numerical results are shown for the critical flow in a LOBI nozzle; the two-phase flow is described by a model with equal phase velocities and thermodynamic non-equilibrium

  19. Analysis of forced convective transient boiling by homogeneous model of two-phase flow

    International Nuclear Information System (INIS)

    Kataoka, Isao

    1985-01-01

    Transient forced convective boiling is of practical importance in relation to the accident analysis of nuclear reactor etc. For large length-to-diameter ratio, the transient boiling characteristics are predicted by transient two-phase flow calculations. Based on homogeneous model of two-phase flow, the transient forced convective boiling for power and flow transients are analysed. Analytical expressions of various parameters of transient two-phase flow have been obtained for several simple cases of power and flow transients. Based on these results, heat flux, velocity and time at transient CHF condition are predicted analytically for step and exponential power increases, and step, exponential and linear velocity decreases. The effects of various parameters on heat flux, velocity and time at transient CHF condition have been clarified. Numerical approach combined with analytical method is proposed for more complicated cases. Solution method for pressure transient are also described. (author)

  20. Analysis of data obtained in two-phase flow tests of primary heat transport pumps

    International Nuclear Information System (INIS)

    Currie, T.C.

    1986-06-01

    This report analyzes data obtained in two-phase flow tests of primary heat transport pumps performed during the period 1980-1983. Phenomena which have been known to cause pump-induced flow oscillations in pressurized piping systems under two-phase conditions are reviewed and the data analyzed to determine whether any of the identified phenomena could have been responsible for the instabilities observed in those tests. Tentative explanations for the most severe instabilities are given based on those analyses. It is shown that suction pipe geometry probably plays an important role in promoting instabilities, so additional experiments to investigate the effect of suction pipe geometry on the stability of flow in a closed pipe loop under two-phase conditions are recommended

  1. Entropy analysis on non-equilibrium two-phase flow models

    Energy Technology Data Exchange (ETDEWEB)

    Karwat, H.; Ruan, Y.Q. [Technische Universitaet Muenchen, Garching (Germany)

    1995-09-01

    A method of entropy analysis according to the second law of thermodynamics is proposed for the assessment of a class of practical non-equilibrium two-phase flow models. Entropy conditions are derived directly from a local instantaneous formulation for an arbitrary control volume of a structural two-phase fluid, which are finally expressed in terms of the averaged thermodynamic independent variables and their time derivatives as well as the boundary conditions for the volume. On the basis of a widely used thermal-hydraulic system code it is demonstrated with practical examples that entropy production rates in control volumes can be numerically quantified by using the data from the output data files. Entropy analysis using the proposed method is useful in identifying some potential problems in two-phase flow models and predictions as well as in studying the effects of some free parameters in closure relationships.

  2. CFD evaluation on the thermohydraulic characteristics of tube support plates in steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, B.; Zhang, H.; Han, B.; Yang, B.W. [Xi' an Jiaotong Univ. (China). School of Nuclear Science and Technology; Mo, S.J.; Ren, H.B.; Qin, J.M.; Zuo, C.P. [China Nuclear Power Design Co. Ltd., ShenZhen (China)

    2016-07-15

    The integrity and thermal hydraulic characteristics of steam generator are of great concern in the nuclear industry. The tube support plates (TSP), one of the most important components of the steam generator, not only support the heat transfer tubes, but also affect the flow dynamic and thermal hydraulic characteristics of the secondary-side flow inside the steam generator. Different working conditions, ranging from single-phase adiabatic condition to two-phase high-void boiling condition, are simulated and analyzed. Calculated void fraction, under simple geometry, agrees well with the experiment data whilst the simulated heat transfer coefficient is tremendously close to the empirical correlation. Temperature, void fraction, and velocity distributions in different locations show reasonable distribution. The simulation results indicate that TSP can enhance the heat transfer in the secondary side of the steam generator. On the top of TSP, with the increase in cross-section flow area, the back-flow phenomenon occurs, which might lead to the contamination of precipitation.

  3. Zero-G two phase flow regime modeling in adiabatic flow

    International Nuclear Information System (INIS)

    Reinarts, T.R.; Best, F.R.; Wheeler, M.; Miller, K.M.

    1993-01-01

    Two-phase flow, thermal management systems are currently being considered as an alternative to conventional, single phase systems for future space missions because of their potential to reduce overall system mass, size, and pumping power requirements. Knowledge of flow regime transitions, heat transfer characteristics, and pressure drop correlations is necessary to design and develop two-phase systems. This work is concerned with microgravity, two-phase flow regime analysis. The data come from a recent sets of experiments. The experiments were funded by NASA Johnson Space Center (JSC) and conducted by NASA JSC with Texas A ampersand M University. The experiment was on loan to NASA JSC from Foster-Miller, Inc., who constructed it with funding from the Air Force Phillips Laboratory. The experiment used R12 as the working fluid. A Foster-Miller two phase pump was used to circulate the two phase mixture and allow separate measurements of the vapor and liquid flow streams. The experimental package was flown 19 times for 577 parabolas aboard the NASA KC-135 aircraft which simulates zero-G conditions by its parabolic flight trajectory. Test conditions included bubbly, slug and annular flow regimes in 0-G. The superficial velocities of liquid and vapor have been obtained from the measured flow rates and are presented along with the observed flow regimes and several flow regime transition predictions. None of the predictions completely describe the transitions as indicated by the data

  4. Experimental modeling of flow-induced vibration of multi-span U-tubes in a CANDU steam generator

    International Nuclear Information System (INIS)

    Mohany, A.; Feenstra, P.; Janzen, V.P.; Richard, R.

    2009-01-01

    Flow-induced vibration of the tubes in a nuclear steam generator is a concern for designers who are trying to increase the life span of these units. The dominant excitation mechanisms are fluidelastic instability and random turbulence excitation. The outermost U-bend region of the tubes is of greatest concern because the flow is almost perpendicular to the tube axis and the unsupported span is relatively long. The support system in this region must be well designed in order to minimize fretting wear of the tubes at the support locations. Much of the previous testing was conducted on straight single-span or cantilevered tubes in cross-flow. However, the dynamic response of steam generator multi-span U-tubes with clearance supports is expected to be different. Accurate modeling of the tube dynamics is important to properly simulate the dynamic interaction of the tube and supports. This paper describes a test program that was developed to measure the dynamic response of a bundle of steam generator U-tubes with Anti-Vibration Bar (AVB) supports, subjected to Freon two-phase cross-flow. The tube bundle has similar geometrical conditions to those expected for future CANDU steam generators. Future steam generators will be larger than previous CANDU steam generators, nearly twice the heat transfer area, with significant changes in process conditions in the U-bend region, such as increased steam quality and a broader range of flow velocities. This test program was initiated at AECL to demonstrate that the tube support design for future CANDU steam generators will meet the stringent requirements associated with a 60 year design life. The main objective of the tests is to address the issue of in-plane and out-of-plane fluidelastic instability and random turbulent excitation of a U-tube bundle with Anti-Vibration Bar (AVB) supports. Details of the test rig, measurement techniques and preliminary instrumentation results are described in the paper. (author)

  5. Evaluation of materials' corrosion and chemistry issues for advanced gas cooled reactor steam generators using full scale plant simulations

    International Nuclear Information System (INIS)

    Woolsey, I.S.; Rudge, A.J.; Vincent, D.J.

    1998-01-01

    Advanced Gas Cooled Reactors (AGRS) employ once-through steam Generators of unique design to provide steam at approximately 530 degrees C and 155 bar to steam turbines of similar design to those of fossil plants. The steam generators are highly compact, and have either a serpentine or helical tube geometry. The tubes are heated on the outside by hot C0 2 gas, and steam is generated on the inside of the tubes. Each individual steam generator tube consists of a carbon steel feed and primary economiser section, a 9%Cr steel secondary economiser, evaporator and primary superheater, and a Type 316L austenitic stainless steel secondary superheater, all within a single tube pass. The multi-material nature of the individual tube passes, the need to maintain specific thermohydraulic conditions within the different material sections, and the difficulties of steam generator inspection and repair, have required extensive corrosion-chemistry test programmes to ensure waterside corrosion does not present a challenge to their integrity. A major part of these programmes has been the use of a full scale steam generator test facility capable of simulating all aspects of the waterside conditions which exist in the plant. This facility has been used to address a wide variety of possible plant drainage/degradation processes. These include; single- and two-phase flow accelerated corrosion of carbon steel, superheat margins requirements and the stress-corrosion behaviour of the austenitic superheaters, on-load corrosion of the evaporator materials, and iron transport and oxide deposition behaviour. The paper outlines a number of these, and indicates how they have been of value in helping to maintain reliable operation of the plant. (author)

  6. Calculations of the nozzle coefficient of discharge of wet steam turbine stages

    International Nuclear Information System (INIS)

    Jinling, Z.; Yinian, C.

    1989-01-01

    A method is presented for calculating the coefficient of discharge of wet steam turbine nozzles. The theoretical formulation of the problem is rigorously in accordance with the theory of two-phase wet steam expansion flow through steam turbine nozzles. The computational values are plotted as sets of curves in accordance with orthogonality test principles. They agree satisfactorily both with historical empirical data and the most recent experimental data obtained in the wet steam two-phase flow laboratory of Xian Jiaotong University. (author)

  7. Void fraction and flow regime determination by optical probe for boiling two-phase flow in a tube subchannel

    International Nuclear Information System (INIS)

    Cheng Huiping; Wu Hongtao; Ba Changxi; Yan Xiaoming; Huang Suyi

    1995-12-01

    In view of the need to determine void fraction and flow regime of vapor-liquid two-phase flow in the steam generator test model, domestic made optical probe was applied on a small-scale freon two-phase flow test rig. Optical probe signals were collected at a sampling rate up to 500 Hz and converted into digital form. Both the time signal, and the amplitude probability density function and FFT spectrum function calculated thereof were analysed in the time and frequency domains respectively. The threshold characterizing vapor or liquid contact with the probe tip was determined from the air-water two-phase flow pressure drop test results. Then, the boiling freon two-phase flow void fraction was determined by single threshold method, and compared with numerical heat transfer computation. Typical patterns which were revealed by the above-mentioned time signal and the functions were found corresponding to distinct flow regimes, as corroborated by visual observation. The experiment shows that the optical probe was a promising technique for two-phase flow void fraction measurement and flow regime identification (3 refs., 15 figs., 1 tab.)

  8. Identification of two-phase flow patterns in a nuclear reactor by the high-frequency contribution fraction

    International Nuclear Information System (INIS)

    Wang, Y.W.; Pei, B.S.; King, C.H.; Lee, S.C.

    1989-01-01

    Recently, King et al. and Wang et al. analyzed the fluctuating characteristics of differential pressure and void fraction by the optimum modeling method and by spectral analysis, respectively. These two investigations presented some new concepts and deterministic criteria, which are based on purely empirical formulas, to identify two-phase flow patterns. These deterministic criteria on two-phase flow patterns' identification seem to show reasonable performance. In King's and Wang's studies, there are at least three problems that need further investigations for the applications to the nuclear reactor engineering field. These three problems are the following: 1. Is the response to a certain two-phase flow pattern, i.e., the fluctuating characteristics, of neutrons the same as that of differential pressure or void fraction? 2. Could those criteria developed from air/water flow be allowed to identify steam/water two-phase flow patterns? 3. Could those criteria be applied to identify two-phase flow patterns in rod bundles? In this paper, parts of the investigated results answer the first problem, and detailed comparisons with the previous work of the authors are given on a variety of items

  9. Steam condensation induced water hammer simulations for different pipelines

    International Nuclear Information System (INIS)

    Barna, I.F.; Ezsol, G.

    2011-01-01

    We investigate steam condensation induced water hammer (CIWH) phenomena and present theoretical results for different kind of pipelines. We analyze the process with the WAHA3 model based on two-phase flow six first-order partial differential equations that present one dimensional, surface averaged mass, momentum and energy balances. A second order accurate high-resolution shock-capturing numerical scheme was applied with different kind of limiters in the numerical calculations. At first, we present calculations for various pipelines in the VVER-440-312 type nuclear reactor. Our recent calculation clearly shows that the six conditions of Griffith are only necessary conditions for CIWH but not sufficient. As second results we performed calculations for various geometries and compare with the theory of Chun. (author)

  10. Simulation of two-phase flows by domain decomposition

    International Nuclear Information System (INIS)

    Dao, T.H.

    2013-01-01

    This thesis deals with numerical simulations of compressible fluid flows by implicit finite volume methods. Firstly, we studied and implemented an implicit version of the Roe scheme for compressible single-phase and two-phase flows. Thanks to Newton method for solving nonlinear systems, our schemes are conservative. Unfortunately, the resolution of nonlinear systems is very expensive. It is therefore essential to use an efficient algorithm to solve these systems. For large size matrices, we often use iterative methods whose convergence depends on the spectrum. We have studied the spectrum of the linear system and proposed a strategy, called Scaling, to improve the condition number of the matrix. Combined with the classical ILU pre-conditioner, our strategy has reduced significantly the GMRES iterations for local systems and the computation time. We also show some satisfactory results for low Mach-number flows using the implicit centered scheme. We then studied and implemented a domain decomposition method for compressible fluid flows. We have proposed a new interface variable which makes the Schur complement method easy to build and allows us to treat diffusion terms. Using GMRES iterative solver rather than Richardson for the interface system also provides a better performance compared to other methods. We can also decompose the computational domain into any number of sub-domains. Moreover, the Scaling strategy for the interface system has improved the condition number of the matrix and reduced the number of GMRES iterations. In comparison with the classical distributed computing, we have shown that our method is more robust and efficient. (author) [fr

  11. Characterizing dynamic hysteresis and fractal statistics of chaotic two-phase flow and application to fuel cells

    International Nuclear Information System (INIS)

    Burkholder, Michael B.; Litster, Shawn

    2016-01-01

    In this study, we analyze the stability of two-phase flow regimes and their transitions using chaotic and fractal statistics, and we report new measurements of dynamic two-phase pressure drop hysteresis that is related to flow regime stability and channel water content. Two-phase flow dynamics are relevant to a variety of real-world systems, and quantifying transient two-phase flow phenomena is important for efficient design. We recorded two-phase (air and water) pressure drops and flow images in a microchannel under both steady and transient conditions. Using Lyapunov exponents and Hurst exponents to characterize the steady-state pressure fluctuations, we develop a new, measurable regime identification criteria based on the dynamic stability of the two-phase pressure signal. We also applied a new experimental technique by continuously cycling the air flow rate to study dynamic hysteresis in two-phase pressure drops, which is separate from steady-state hysteresis and can be used to understand two-phase flow development time scales. Using recorded images of the two-phase flow, we show that the capacitive dynamic hysteresis is related to channel water content and flow regime stability. The mixed-wettability microchannel and in-channel water introduction used in this study simulate a polymer electrolyte fuel cell cathode air flow channel.

  12. Characterizing dynamic hysteresis and fractal statistics of chaotic two-phase flow and application to fuel cells

    Energy Technology Data Exchange (ETDEWEB)

    Burkholder, Michael B.; Litster, Shawn, E-mail: litster@andrew.cmu.edu [Department of Mechanical Engineering, Carnegie Mellon University, Pittsburgh, Pennsylvania 15213 (United States)

    2016-05-15

    In this study, we analyze the stability of two-phase flow regimes and their transitions using chaotic and fractal statistics, and we report new measurements of dynamic two-phase pressure drop hysteresis that is related to flow regime stability and channel water content. Two-phase flow dynamics are relevant to a variety of real-world systems, and quantifying transient two-phase flow phenomena is important for efficient design. We recorded two-phase (air and water) pressure drops and flow images in a microchannel under both steady and transient conditions. Using Lyapunov exponents and Hurst exponents to characterize the steady-state pressure fluctuations, we develop a new, measurable regime identification criteria based on the dynamic stability of the two-phase pressure signal. We also applied a new experimental technique by continuously cycling the air flow rate to study dynamic hysteresis in two-phase pressure drops, which is separate from steady-state hysteresis and can be used to understand two-phase flow development time scales. Using recorded images of the two-phase flow, we show that the capacitive dynamic hysteresis is related to channel water content and flow regime stability. The mixed-wettability microchannel and in-channel water introduction used in this study simulate a polymer electrolyte fuel cell cathode air flow channel.

  13. Contribution to the theory of the two phase blowdown phenomenon

    International Nuclear Information System (INIS)

    Hutcherson, M.N.

    1975-12-01

    In order to accurately model the two phase portion of a pressure vessel blowdown, it becomes necessary to understand the bubble growth mechanism within the vessel during the early period of the decompression, the two phase flow behavior within the vessel, and the applicability of the available two phase critical flow models to the blowdown transient. To aid in providing answers to such questions, a small scale, separate effects, isothermal blowdown experiment has been conducted in a small pressure vessel. The tests simulated a full open, double ended, guillotine break in a large diameter, short exhaust duct from the vessel. The vaporization process at the initiation of the decompression is apparently that of thermally dominated bubble growth originating from the surface cavities inside the system. Thermodynamic equilibrium of the remaining fluid within the vessel existed in the latter portion of the decompression. A nonuniform distribution of fluid quality within the vessel was also detected in this experiment. By comparison of the experimental results from this and other similar transient, two phase critical flow studies with steady state, small duct, two phase critical flow data, it is shown that transient, two phase critical flow in large ducts appears to be similar to steady state, two phase critical flow in small ducts. Analytical models have been developed to predict the blowdown characteristics of a system during subcooled decompression, the bubble growth regime of blowdown, and also in the nearly dispersed period of depressurization. This analysis indicates that the system pressure history early in the blowdown is dependent on the internal vessel surface area, the internal vessel volume, and also on the exhaust flow area from the system. This analysis also illustrates that the later period of decompression can be predicted based on thermodynamic equilibrium

  14. Switching moving boundary models for two-phase flow evaporators and condensers

    Science.gov (United States)

    Bonilla, Javier; Dormido, Sebastián; Cellier, François E.

    2015-03-01

    The moving boundary method is an appealing approach for the design, testing and validation of advanced control schemes for evaporators and condensers. When it comes to advanced control strategies, not only accurate but fast dynamic models are required. Moving boundary models are fast low-order dynamic models, and they can describe the dynamic behavior with high accuracy. This paper presents a mathematical formulation based on physical principles for two-phase flow moving boundary evaporator and condenser models which support dynamic switching between all possible flow configurations. The models were implemented in a library using the equation-based object-oriented Modelica language. Several integrity tests in steady-state and transient predictions together with stability tests verified the models. Experimental data from a direct steam generation parabolic-trough solar thermal power plant is used to validate and compare the developed moving boundary models against finite volume models.

  15. Damping and fluidelastic instability in two-phase cross-flow heat exchanger tube arrays

    Science.gov (United States)

    Moran, Joaquin E.

    An experimental study was conducted to investigate damping and fluidelastic instability in tube arrays subjected to two-phase cross-flow. The purpose of this research was to improve our understanding of these phenomena and how they are affected by void fraction and flow regime. The model tube bundle had 10 cantilevered tubes in a parallel-triangular configuration, with a pitch ratio of 1.49. The two-phase flow loop used in this research utilized Refrigerant 11 as the working fluid, which better models steam-water than air-water mixtures in terms of vapour-liquid mass ratio as well as permitting phase changes due to pressure fluctuations. The void fraction was measured using a gamma densitometer, introducing an improvement over the Homogeneous Equilibrium Model (HEM) in terms of void fraction, density and velocity predictions. Three different damping measurement methodologies were implemented and compared in order to obtain a more reliable damping estimate. The methods were the traditionally used half-power bandwidth, the logarithmic decrement and an exponential fitting to the tube decay response. The decay trace was obtained by "plucking" the monitored tube from outside the test section using a novel technique, in which a pair of electromagnets changed their polarity at the natural frequency of the tube to produce resonance. The experiments showed that the half-power bandwidth produces higher damping values than the other two methods. The primary difference between the methods is caused by tube frequency shifting, triggered by fluctuations in the added mass and coupling between the tubes, which depend on void fraction and flow regime. The exponential fitting proved to be the more consistent and reliable approach to estimating damping. In order to examine the relationship between the damping ratio and mass flux, the former was plotted as a function of void fraction and pitch mass flux in an iso-contour plot. The results showed that damping is not independent of mass

  16. Thermal-hydraulic tests of steam-generator tube-support-plate crevices. Volume 2. Appendixes I through S. Final report

    International Nuclear Information System (INIS)

    Cassell, D.S.; Vroom, D.W.

    1983-01-01

    A test program was conducted to determine for selected steam generator tube supports the thermal/hydraulic conditions at the inception of dryout as indicated by a tube wall temperature excursion, to determine the pressure drop across the supports, and to obtain photographic documentation of the flow upstream and downstream of the supports. A multi-tube steam generator model was used and testing performed over the range of typcal PWR steam generator operating conditions. These appendices contain information on instrumentation calibration, test model and loop calibration, error analysis, test model thermal-hydraulic analyses, index of lab materials and log sheets, index of two-phase flow still photographs, index of high speed movies and video, test data printouts, test model and loop fabrication drawings, procedure for silver brazing tubewall thermocouples, and procedure for esablishing tube-tube support line contact

  17. Behavior of sulfur species in steam generator conditions of PWRs - towards an update of the secondary side corrosion cracking model based on laboratory tests in sulfate environments

    International Nuclear Information System (INIS)

    Mansour, C.; Legras, L.; Catalette, H.; Lefevre, G.; Fedoroff, M.; Pavageau, E.-M.

    2007-01-01

    Secondary side corrosion cracking affects Mill Annealed Alloy 600 steam generator (SG) tubes of PWRs in flow restricted areas where pollutants, such as sulfate, can concentrate and form various aggressive local environments. The 'sulfate model', based on laboratory tests in sulfate environments, was developed to predict the degradation of SG tubes. Such prediction is aimed to be done after having evaluated the chemistry in the flow restricted areas where the degradation occurs. For such purpose, a better knowledge of the behavior of sulfur species in SG conditions is needed. After a brief description of the sulfate model, this paper focuses on the latest experimental results that have been obtained : sorption of sulfur species over magnetite in SG temperature conditions, thermodynamical calculations as well as transmission electron microscopy and X-ray photoelectron spectroscopy observations of magnetite after sorption and C-Ring specimens tested in sulfate environments. Then, all these results are discussed in order to contribute to a better understanding of the secondary side corrosion cracking of SG tubes. (author)

  18. Fluid-elastic vibration in two-phase cross flow

    International Nuclear Information System (INIS)

    Sasakawa, T.; Serizawa, A.; Kawara, Z.

    2003-01-01

    The present work aims at clarifying the mechanisms of fluid elastic vibration of tube bundles in two-phase cross flow. The experiment is conducted using air-water two-phase flow under atmospheric pressure. The test section is a 1.03m long transparent acrylic square duct with 128 x 128 mm 2 cross section, which consists of 3 rod-rows with 5 rods in each row. The rods are 125mm long aluminum rods with 22 mm in diameter (p/D=1.45). The natural frequency of rod vibration is about 30Hz. The result indicated a diversion of observed trend in vibration behavior depending on two-phase flow patterns either bubbly flow or churn flow. Specifically, in churn flow, the fluid elastic vibration has been observed to occur when the frequency in void fraction fluctuation approached to the natural frequency of the rods, but this was not the case in fluid elastic vibration in bubbly flow. This fact suggests the existence of mechanisms closely coupled with two-phase flow structures depending on the flow patterns, that is, static two-phase character-controlled mechanism in bubbly flow and dynamic character- controlled in churn flow

  19. Two-phase flow induced parametric vibrations in structural systems

    International Nuclear Information System (INIS)

    Hara, Fumio

    1980-01-01

    This paper is divided into two parts concerning piping systems and a nuclear fuel pin system. The significant experimental results concerning the random vibration induced in an L-shaped pipe by air-water two-phase flow and the theoretical analysis of the vibration are described in the first part. It was clarified for the first time that the parametric excitation due to the periodic changes of system mass, centrifugal force and Coriolis force was the mechanism of exciting the vibration. Moreover, the experimental and theoretical analyses of the mechanism of exciting vibration by air-water two-phase flow in a straight, horizontal pipe were carried out, and the first natural frequency of the piping system was strongly related to the dominant frequency of void signals. The experimental results on the vibration of a nuclear fuel pin model in parallel air-water two-phase flow are reported in the latter part. The relations between vibrational strain variance and two-phase flow velocity or pressure fluctuation, and the frequency characteristics of vibrational strain variance were obtained. The theoretical analysis of the dynamic interaction between air-water two-phase flow and a fuel pin structure, and the vibrational instability of fuel pins in alternate air and water slugs or in large bubble flow are also reported. (Kako, I.)

  20. Determination of the onset nonlinearity hydrodynamic characteristics at two-phase flow in parallel vertical channels

    International Nuclear Information System (INIS)

    Jovic, V.; Afgan, N.; Jovic, L.; Spasojevic, D.

    1993-01-01

    The paper presents results of the experimental and theoretical analyses of linear and nonlinear characteristics of adiabatic two-phase water-air flow in vertical parallel channels. Regime character changes and linear to nonlinear dynamic characteristics transfer conditions were defined. (author)

  1. Analyses of liquid-gas two-phase flow in fermentation tanks

    International Nuclear Information System (INIS)

    Toi, Takashi; Serizawa, Akimi; Takahashi, Osamu; Kawara, Zensaku; Gofuku, Akio; Kataoka, Isao.

    1993-01-01

    The understanding of two-phase flow is one of the important problems for both design and safety analyses of various engineering systems. For example, the flow conditions in beer fermentation tanks have an influence on the quality of production and productivity of tank. In this study, a two-dimensional numerical calculation code based on the one-pressure two-fluid model is developed to understand the circulation structure of low quality liquid-gas two-phase flows induced by bubble plume in a tank. (author)

  2. Two-phase coolant pump model of pressurized light water nuclear reactors

    International Nuclear Information System (INIS)

    Santos, G.A. dos; Freitas, R.L.

    1990-01-01

    The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The homologous curves set up the complete performance of the pump and are input for accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

  3. Analysis of two-phase flow induced vibrations in perpendiculary supported U-type piping systems

    International Nuclear Information System (INIS)

    Hiramatsu, Tsutomu; Komura, Yoshiaki; Ito, Atsushi.

    1984-01-01

    The perpose of this analysis is to predict the vibration level of a pipe conveying a two-phase flowing fluid. Experiments were carried out with a perpendiculary supported U-type piping system, conveying an air-water two-phase flow in a steady state condition. Fluctuation signals are observed by a void signal sensor, and power spectral densities and probability density functions are obtained from the void signals. Theoretical studies using FEM and an estimation of the exciting forces from the PSD of void signals, provided a good predictional estimation of vibration responses of the piping system. (author)

  4. Manual for operation of the multipurpose thermalhydraulic test facility TOPFLOW (Transient Two Phase Flow Test Facility)

    International Nuclear Information System (INIS)

    Beyer, M.; Carl, H.; Schuetz, H.; Pietruske, H.; Lenk, S.

    2004-07-01

    The Forschungszentrum Rossendorf (FZR) e. V. is constructing a new large-scale test facility, TOPFLOW, for thermalhydraulic single effect tests. The acronym stands for transient two phase flow test facility. It will mainly be used for the investigation of generic and applied steady state and transient two phase flow phenomena and the development and validation of models of computational fluid dynamic (CFD) codes. The manual of the test facility must always be available for the staff in the control room and is restricted condition during operation of personnel and also reconstruction of the facility. (orig./GL)

  5. Development of a large-scale general purpose two-phase flow analysis code

    International Nuclear Information System (INIS)

    Terasaka, Haruo; Shimizu, Sensuke

    2001-01-01

    A general purpose three-dimensional two-phase flow analysis code has been developed for solving large-scale problems in industrial fields. The code uses a two-fluid model to describe the conservation equations for two-phase flow in order to be applicable to various phenomena. Complicated geometrical conditions are modeled by FAVOR method in structured grid systems, and the discretization equations are solved by a modified SIMPLEST scheme. To reduce computing time a matrix solver for the pressure correction equation is parallelized with OpenMP. Results of numerical examples show that the accurate solutions can be obtained efficiently and stably. (author)

  6. NOTICONA--a nonlinear time-domain computer code of two-phase natural circulation instability

    International Nuclear Information System (INIS)

    Su Guanghui; Guo Yujun; Zhang Jinling; Qiu Shuizheng; Jia Dounan; Yu Zhenwan

    1997-10-01

    A microcomputer code, NOTICONA, is developed, which is used for non-linear analysing the two-phase natural circulation systems. The mathematical model of the code includes point source neutron-kinetic model, the feedback of reactivity model, single-phase and two-phase flow model, heat transfer model in different conditions, associated model, etc. NOTICONA is compared with experiments, and its correctness and accuracy are proved. Using NOTICONA, the density wave oscillation (type I) of the 5 MW Test Heating Reactor are calculated, and the marginal stability boundary is obtained

  7. Two-phase flow void fraction measurement using gamma ray attenuation technique

    International Nuclear Information System (INIS)

    Silva, R.D. da.

    1985-01-01

    The present work deals with experimental void fraction measurements in two-phase water-nitrogen flow, by using a gamma ray attenuation technique. Several upward two-phase flow regimes in a vertical tube were simulated. The water flow was varied from 0.13 to 0.44 m 3 /h while the nitrogen flow was varied between 0.01 and 0.1 m 3 /h. The mean volumetric void fraction was determined based on the measured linear void fraction for each flow condition. The results were compared with other authors data and showed a good agreement. (author) [pt

  8. Complex network analysis in inclined oil–water two-phase flow

    International Nuclear Information System (INIS)

    Zhong-Ke, Gao; Ning-De, Jin

    2009-01-01

    Complex networks have established themselves in recent years as being particularly suitable and flexible for representing and modelling many complex natural and artificial systems. Oil–water two-phase flow is one of the most complex systems. In this paper, we use complex networks to study the inclined oil–water two-phase flow. Two different complex network construction methods are proposed to build two types of networks, i.e. the flow pattern complex network (FPCN) and fluid dynamic complex network (FDCN). Through detecting the community structure of FPCN by the community-detection algorithm based on K-means clustering, useful and interesting results are found which can be used for identifying three inclined oil–water flow patterns. To investigate the dynamic characteristics of the inclined oil–water two-phase flow, we construct 48 FDCNs under different flow conditions, and find that the power-law exponent and the network information entropy, which are sensitive to the flow pattern transition, can both characterize the nonlinear dynamics of the inclined oil–water two-phase flow. In this paper, from a new perspective, we not only introduce a complex network theory into the study of the oil–water two-phase flow but also indicate that the complex network may be a powerful tool for exploring nonlinear time series in practice. (general)

  9. Application of two-phase flow for cooling of hybrid microchannel PV cells: A comparative study

    International Nuclear Information System (INIS)

    Valeh-e-Sheyda, Peyvand; Rahimi, Masoud; Karimi, Ebrahim; Asadi, Masomeh

    2013-01-01

    Highlights: ► Showing cooling potential of gas–liquid two-phase flow in microchannels for PV cell. ► Introducing the concept of using slug flow in microchannels for cooling of PV cells. ► In single-phase flow, increasing the liquid flow rate enhances the PV power. ► Showing that in two-phase flow the output power related the fluid flow regime. ► By coupling PV and microchannel an increase up to 38% in output power was observed. - Abstract: This paper reports the experimental data from performance of two-phase flows in a small hybrid microchannel solar cell. Using air and water as two-phase fluid, the experiments were conducted at indoor condition in an array of rectangular microchannels with a hydraulic diameter of 0.667 mm. The gas superficial velocity ranges were between 0 and 3.27 m s −1 while liquid flow rate was 0.04 m s −1 . The performance analysis of the PV cell at slug and transitional slug/annular flow regimes are the focus of this study. The influence of two-phase working fluid on PV cell cooling was compared with single-phase. In addition, the great potential of slug flow for heat removal enhancement in PV/T panel was investigated. The obtained data showed the proposed hybrid system could substantially increases the output power of PV solar cells

  10. Computational fluid dynamics modeling of two-phase flow in a BWR fuel assembly. Final CRADA Report

    International Nuclear Information System (INIS)

    Tentner, A.

    2009-01-01

    A direct numerical simulation capability for two-phase flows with heat transfer in complex geometries can considerably reduce the hardware development cycle, facilitate the optimization and reduce the costs of testing of various industrial facilities, such as nuclear power plants, steam generators, steam condensers, liquid cooling systems, heat exchangers, distillers, and boilers. Specifically, the phenomena occurring in a two-phase coolant flow in a BWR (Boiling Water Reactor) fuel assembly include coolant phase changes and multiple flow regimes which directly influence the coolant interaction with fuel assembly and, ultimately, the reactor performance. Traditionally, the best analysis tools for this purpose of two-phase flow phenomena inside the BWR fuel assembly have been the sub-channel codes. However, the resolution of these codes is too coarse for analyzing the detailed intra-assembly flow patterns, such as flow around a spacer element. Advanced CFD (Computational Fluid Dynamics) codes provide a potential for detailed 3D simulations of coolant flow inside a fuel assembly, including flow around a spacer element using more fundamental physical models of flow regimes and phase interactions than sub-channel codes. Such models can extend the code applicability to a wider range of situations, which is highly important for increasing the efficiency and to prevent accidents.

  11. Three layer model analysis on two-phase critical flow through a converging nozzle

    International Nuclear Information System (INIS)

    Ochi, J.; Ayukawa, K.

    1991-01-01

    A three layer model is proposed for a two-phase critical flow through a converging nozzle in this paper. Most previous analyses of the two phase flow have been based on a homogeneous or a separated flow model as the conservation equations. These results were found to have large deviations from the actual measurements for two phase critical flows. The presented model is based on the assumption that a flow consists of three layers with a mixing region between gas and liquid phase layers. The effect of gas and liquid fraction occupied in the mixing layer was made clear from the numerical results. The measurements of the critical flow rate and the pressure profiles through a converging nozzle were made with air-water flow. The calculated results of these models are discussed in comparison with the experimental data for the flow rates and the pressure distributions under critical conditions

  12. CANDU steam generator life management

    International Nuclear Information System (INIS)

    Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

    1998-01-01

    Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. The research and development program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factory. (author)

  13. Numerical flow analyses of a two-phase hydraulic coupling

    Energy Technology Data Exchange (ETDEWEB)

    Hur, N.; Kwak, M.; Moshfeghi, M. [Sogang University, Seoul (Korea, Republic of); Chang, C.-S.; Kang, N.-W. [VS Engineering, Seoul (Korea, Republic of)

    2017-05-15

    We investigated flow characteristics in a hydraulic coupling at different charged water conditions and speed ratios. Hence, simulations were performed for three-dimensional two-phase flow by using the VOF method. The realizable k-ε turbulence model was adopted. To resolve the interaction of passing blades of the primary and secondary wheels, simulations were conducted in the unsteady framework using a sliding grid technique. The results show that the water-air distribution inside the wheel is strongly dependent upon both amount of charged water and speed ratio. Generally, air is accumulated in the center of the wheel, forming a toroidal shape wrapped by the circulating water. The results also show that at high speed ratios, the solid-body-like rotation causes dry areas on the periphery of the wheels and, hence, considerably decreases the circulating flow rate and the transmitted torque. Furthermore, the momentum transfer was investigated through the concept of a mass flux triangle based on the local velocity multiplied by the local mixture density instead of the velocity triangle commonly used in a single-phase turbomachine analysis. Also, the mass fluxes along the radius of the coupling in the partially charged and fully charged cases were found to be completely different. It is shown that the flow rate at the interfacial plane and also the transmitted torque are closely related and are strongly dependent upon both the amount of charged water and speed ratio. Finally, a conceptual categorization together with two comprehensive maps was provided for the torque transmission and also circulating flow rates. These two maps in turn exhibit valuable engineering information and can serve as bases for an optimal design of a hydraulic coupling.

  14. Abnormal breakdown characteristic in a two-phase mixture

    Energy Technology Data Exchange (ETDEWEB)

    Ye Qizheng; Li Jin; Lu Fei [College of Electrical and Electronic Engineering, Huazhong University of Science and Technology, Wuhan, 430074 (China)

    2006-05-21

    A two-phase mixture (TPM) is a mixture of gas and macroparticles of high concentration. Based on Townsend's theory, a new cell-iterative model in analytical form for the breakdown mechanism in TPM is presented. Compared with the original cell-iterative model in our previous paper, the obstructive factor of the macroparticles that influences the electron avalanche propagation is considered, except for the macroparticles distorting the electrical field and capture of the electrons. The cell attractive parameter k is presented according to the classical continuum theory for field charging. The modified Paschen law for a TPM is presented to calculate the breakdown voltage. The breakdown voltage of the TPM, U{sub TPM}, increases gradually with an increase in the macroparticle number density (m). The voltage U{sub TPM} is lower than that of the pure gas at low m values and larger at high m values. With a decrease of the macroparticle volume fraction and the dielectric mismatch, the voltage U{sub TPM} increases gradually at low m values and decreases gradually at high m values. The voltage U{sub TPM} at pd 200 cm Torr is lower than that at pd = 760 cm Torr for low m values and larger for high m values. This kind of abnormal breakdown characteristic in the TPM occurs in the case of high macroparticle volume fraction. On the other hand, the minimum of the TPM's Paschen curve increases with increase in m. It provides the possibility and the conditions of greatly increasing the breakdown voltage in a nearly uniform field.

  15. Two-Phase Phenomena In Wet Flue Gas Desulfurization Process

    International Nuclear Information System (INIS)

    Minzer, U.; Moses, E.J.; Toren, M.; Blumenfeld, Y.

    1998-01-01

    In order to reduce sulfur oxides discharge, Israel Electric Corporation (IEC) is building a wet Flue Gas Desulfurization (FGD) facility at Rutenberg B power station. The primary objective of IEC is to minimize the occurrence of stack liquid discharge and avoid the discharge of large droplets, in order to prevent acid rain around the stack. Liquid discharge from the stack is the integrated outcome of two-phase processes, which are discussed in this work. In order to estimate droplets discharge the present investigation employs analytical models, empirical tests, and numerical calculations of two-phase phenomena. The two-phase phenomena are coupled and therefore cannot be investigated separately. The present work concerns the application of Computational Fluid Dynamic (CFD) as an engineering complementary tool in the IEC investigation

  16. DISTRIBUTION OF TWO-PHASE FLOW IN A DISTRIBUTOR

    Directory of Open Access Journals (Sweden)

    AZRIDJAL AZIZ

    2012-02-01

    Full Text Available The flow configuration and distribution behavior of two-phase flow in a distributor made of acrylic resin have been investigated experimentally. In this study, air and water were used as two-phase flow working fluids. The distributor consists of one inlet and two outlets, which are set as upper and lower, respectively. The flow visualization at the distributor was made by using a high–speed camera. The flow rates of air and water flowing out from the upper and lower outlet branches were measured. Effects of inclination angle of the distributor were investigated. By changing the inclination angle from vertical to horizontal, uneven distributions were also observed. The distribution of two-phase flow through distributor tends even flow distribution on the vertical position and tends uneven distribution on inclined and horizontal positions. It is shown that even distribution could be achieved at high superficial velocities of both air and water.

  17. Mathematical modeling of disperse two-phase flows

    CERN Document Server

    Morel, Christophe

    2015-01-01

    This book develops the theoretical foundations of disperse two-phase flows, which are characterized by the existence of bubbles, droplets or solid particles finely dispersed in a carrier fluid, which can be a liquid or a gas. Chapters clarify many difficult subjects, including modeling of the interfacial area concentration. Basic knowledge of the subjects treated in this book is essential to practitioners of Computational Fluid Dynamics for two-phase flows in a variety of industrial and environmental settings. The author provides a complete derivation of the basic equations, followed by more advanced subjects like turbulence equations for the two phases (continuous and disperse) and multi-size particulate flow modeling. As well as theoretical material, readers will discover chapters concerned with closure relations and numerical issues. Many physical models are presented, covering key subjects including heat and mass transfers between phases, interfacial forces and fluid particles coalescence and breakup, a...

  18. Development and calibration of instruments for measurements in transient two-phase flow

    International Nuclear Information System (INIS)

    Banerjee, S.; Heidrick, T.R.

    1981-01-01

    For validation and development of theoretical models for transient two-phase flow, it is necessary to measure local and cross-sectionally averaged thermalhydraulic parameters. Of these parameters, void fraction and mass velocity are the most difficult to measure. In this paper, we present our recent work on various techniques for determining these quantities. The possibility of determining flow regime by using fast neutron transmission is discussed. The development of a miniaturized electrical resistivity probe for measuring local void fraction is described, together with calibrations obtained by integrating the void fraction profile and comparing the cross-sectionally averaged void fraction with direct measurements using two quick closing valves. Results on the calibration of combinations of full-flow turbine meters, Pitot tube rakes and gamma densitometers for measuring cross-sectionally averaged mass velocity in steady steam-water flow are presented. The results are interpreted with a simple model using single-phase calibration factors for the Pitot tube rakes and turbine meters. Calibration experiments were also done in transient steam-water flows and interpretation of the results with the steady state models is also discussed

  19. Numerical simulation of gas-liquid two-phase flow behavior with condensation heat transfer

    International Nuclear Information System (INIS)

    Takamori, Kazuhide; Murase, Michio; Baba, Yoshikazu; Aihara, Tsuyoshi.

    1995-01-01

    In this study, condensation heat transfer experiments were performed in order to verify a condensation heat transfer model coupled with a three-dimensional two-phase flow analysis. In the heat transfer model, the liquid film flow rate on the heat transfer tubes was calculated by a mass balance equation and the liquid film thickness was calculated from the liquid film flow rate using Nusselt's laminar flow model and Fujii's equation for steam velocity effect. In the experiments, 112 horizontal staggered tubes with an outer diameter of 16 mm and length of 0.55 m were used. Steam and spray water were supplied to the test section, and inlet quality was controlled by the spray water flow rate. The temperature was 100degC and the pressure was 0.1 MPa. The overall heat transfer coefficients were measured for inlet quality of 13-100%. From parameter calculations for the falling liquid film ratio from the upper tubes to the lower tubes, the calculated overall heat transfer coefficients agreed with the data to within ±5% at the falling liquid film ratio of 0.7. (author)

  20. Behavior of pumps conveying two-phase liquid flow

    International Nuclear Information System (INIS)

    Grison, Pierre; Lauro, J.-F.

    1979-01-01

    Determination of the two-phase flow (critical or otherwise) through a pump is an essential requirement for complete description of a loss of primary coolant accident in a PWR plant. Theoretical and experimental research at Electricite de France on this subject is described and problems associated with the introduction of a two-phase fluid (with mass transfer) are discussed, with an attempt to single out new phenomena involved and establish their effect on pump behavior. A complementary experimental investigation is described and the results of tests at pressures and temperatures up to 120 bars and 320 0 C respectively are compared with the theoretical model data [fr

  1. Behavior of pumps conveying two-phase liquid flow

    Energy Technology Data Exchange (ETDEWEB)

    Grison, P; Lauro, J F [Electricite de France, 78 - Chatou. Direction des Etudes et Recherches

    1979-01-01

    Determination of the two-phase flow (critical or otherwise) through a pump is an essential requirement for complete description of a loss of primary coolant accident in a PWR plant. Theoretical and experimental research at Electricite de France on this subject is described and problems associated with the introduction of a two-phase fluid (with mass transfer) are discussed, with an attempt to single out new phenomena involved and establish their effect on pump behavior. A complementary experimental investigation is described and the results of tests at pressures and temperatures up to 120 bars and 320/sup 0/C respectively are compared with the theoretical model data.

  2. Dynamic Modeling of Phase Crossings in Two-Phase Flow

    DEFF Research Database (Denmark)

    Madsen, Søren; Veje, Christian; Willatzen, Morten

    2012-01-01

    by a high resolution finite difference scheme due to Kurganov and Tadmore. The homogeneous formulation requires a set of thermodynamic relations to cover the entire range from liquid to gas state. This leads a number of numerical challenges since these relations introduce discontinuities in the derivative...... of the variables and are usually very slow to evaluate. To overcome these challenges, we use an interpolation scheme with local refinement. The simulations show that the method handles crossing of the saturation lines for both liquid to two-phase and two-phase to gas regions. Furthermore, a novel result obtained...

  3. Two-phase LMMHD mixer-development experiments

    International Nuclear Information System (INIS)

    Fabris, G.; Dunn, P.F.; Chow, J.C.F.

    1978-01-01

    The results of a series of experiments conducted to evaluate the fluid mechanical performance of various two-phase LMMHD mixer designs are presented. The results from both flow visualization studies of the local two-phase flows downstream from various mixer-element configurations and local measurements performed to characterize these flows are presented. A conceptual LMMHD mixer design is described that insures the generation of small bubbles, prevents the formation of gas slugs and separated regions, and favors the stabilization of a homogeneous foam flow

  4. Two-phase flow model with nonequilibrium and critical flow

    International Nuclear Information System (INIS)

    Sureau, H.; Houdayer, G.

    1976-01-01

    The model proposed includes the three conservation equations (mass, momentum, energy) applied to the two phase flows and a fourth partial derivative equation which takes into account the nonequilibriums and describes the mass transfer process. With this model, the two phase critical flow tests performed on the Moby-Dick loop (CENG) with several geometries, are interpreted by a unique law. Extrapolations to industrial dimension problems show that geometry and size effects are different from those obtained with earlier models (Zaloudek, Moody, Fauske) [fr

  5. A void fraction model for annular two-phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Tandon, T.N.; Gupta, C.P.; Varma, H.K.

    1985-01-01

    An analytical model has been developed for predicting void fraction in two-phase annular flow. In the analysis, the Lockhart-Martinelli method has been used to calculate two-phase frictional pressure drop and von Karman's universal velocity profile is used to represent the velocity distribution in the annular liquid film. Void fractions predicted by the proposed model are generally in good agreement with a available experimental data. This model appears to be as good as Smith's correlation and better than the Wallis and Zivi correlations for computing void fraction.

  6. Comparison of Experimental and Numerical Two-Phase Flows in a Porous Micro-Model

    Directory of Open Access Journals (Sweden)

    Dustin Crandall

    2009-12-01

    Full Text Available Characterizing two-phase flow in porous media is important to provide estimates of sweep efficiency in enhanced oil recovery and storage estimates in potential geological CO2 sequestration repositories. To further the current understanding of two-phase flow in porous media a micro-model of interconnected channels was designed and fabricated using stereolithography to experimentally study gas-liquid flows. This flowcell was created with a wide variability of throat dimensions to represent naturally occurring porous media. Low flow rate experiments of immiscible two-phase drainage were performed within this cell. Additionally, a computational model for analyzing two-phase flows in the same flowcell was developed and used to simulate conditions not possible with our laboratory settings. The computational model was first tested for the identical conditions used in the experimental studies, and was shown to be in good agreement with the experimentally determined fractal dimension of the invading gas structure, time until breakthrough, and fluid saturation. The numerical model was then used to study two-phase air-water flows in flowcells with the same geometry and different gas-liquid-solid contact angles. The percent saturation of air and the motion of the fluids through the cell were found to vary with changes in these parameters. Finally, to simulate flows expected during geologic carbon sequestration, the fluid properties and interface conditions were set to model the flow of CO2 into a brine-saturated porous medium at representative subsurface conditions. The CO2 flows were shown to have larger gas saturations than the previous air into water studies. Thus the accuracy of the computational model was supported by the flowcell experiments, and the computational model extended the laboratory results to conditions not possible with the apparatus used in the experiments.

  7. Regulation of ageing steam generators

    International Nuclear Information System (INIS)

    Jarman, B.L.; Grant, I.M.; Garg, R.

    1998-01-01

    Recent years have seen leaks and shutdowns of Canadian CANDU plants due to steam generator tube degradation by mechanisms including stress corrosion cracking, fretting and pitting. Failure of a single steam generator tube, or even a few tubes, would not be a serious safety related event in a CANDU reactor. The leakage from a ruptured tube is within the makeup capacity of the primary heat transport system, so that as long as the operator takes the correct actions, the off-site consequences will be negligible. However, assurance that no tubes deteriorate to the point where their integrity could be seriously breached as result of potential accidents, and that any leakage caused by such an accident will be small enough to be inconsequential, can only be obtained through detailed monitoring and management of steam generator condition. This paper presents the AECB's current approach and future regulatory directions regarding ageing steam generators. (author)

  8. Inlet effects on vertical-downward air–water two-phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Qiao, Shouxu; Mena, Daniel; Kim, Seungjin, E-mail: skim@psu.edu

    2017-02-15

    Highlights: • Inlet effects on two-phase flow parameters in vertical-downward flow are studied. • Flow regimes in the vertical-downward two-phase flow are defined. • Vertical-downward flow regime maps for three inlet configurations are developed. • Frictional pressure loss analysis for three different inlets is performed. • Database of local two-phase flow parameters for each inlet configuration. - Abstract: This paper focuses on investigating the geometric effects of inlets on global and local two-phase flow parameters in vertical-downward air–water two-phase flow. Flow visualization, frictional pressure loss analysis, and local experiments are performed in a test facility constructed from 50.8 mm inner diameter acrylic pipes. Three types of inlets of interest are studied: (1) two-phase flow injector without a flow straightener (Type A), (2) two-phase flow injector with a flow straightener (Type B), and (3) injection through a horizontal-to-vertical-downward 90° vertical elbow (Type C). A detailed flow visualization study is performed to characterize flow regimes including bubbly, slug, churn-turbulent, and annular flow. Flow regime maps for each inlet are developed and compared to identify the effects of each inlet. Frictional pressure loss analysis shows that the Lockhart–Martinelli method is capable of correlating the frictional loss data acquired for Type B and Type C inlets with a coefficient value of C = 25, but additional data may be needed to model the Type A inlet. Local two-phase flow parameters measured by a four-sensor conductivity probe in four bubbly and near bubbly flow conditions are analyzed. It is observed that vertical-downward two-phase flow has a characteristic center-peaked void profile as opposed to a wall-peaked profile as seen in vertical-upward flow. Furthermore, it is shown that the Type A inlet results in the most pronounced center-peaked void fraction profile, due to the coring phenomenon. Type B and Type C inlets

  9. Development of two phase turbulent mixing model for subchannel analysis relevant to BWR

    International Nuclear Information System (INIS)

    Sharma, M.P.; Nayak, A.K.; Kannan, Umasankari

    2014-01-01

    A two phase flow model is presented, which predicts both liquid and gas phase turbulent mixing rate between adjacent subchannels of reactor rod bundles. The model presented here is for slug churn flow regime, which is dominant as compared to the other regimes like bubbly flow and annular flow regimes, since turbulent mixing rate is the highest in slug churn flow regime. In this paper, we have defined new dimensionless parameters i.e. liquid mixing number and gas mixing number for two phase turbulent mixing. The liquid mixing number is a function of mixture Reynolds number whereas the gas phase mixing number is a function of both mixture Reynolds number and volumetric fraction of gas. The effect of pressure, geometrical influence of subchannel is also included in this model. The present model has been tested against low pressure and temperature air-water and high pressure and temperature steam-water experimental data found that it shows good agreement with available experimental data. (author)

  10. IDRIFF two-phase simulation code and its application to the study of a pressurizer

    International Nuclear Information System (INIS)

    Sollychin, R.; Garland, W.J.; Chang, J.S.

    1987-01-01

    The simulation code IDRIFF (Integrated Drift-flux Formulation) has been developed as a convenient tool in two-phase flow analysis, which demands the following two conflicting requirements: (a) provision for detailed information on local phenomena in the flow;(b) fast calculation of averaged values of parameters for engineering type flow problems. A small scale pressurizer made of a glass tank and its associated systems were set-up to simulate the behavior of nuclear power plant pressurizer. Flow-pattern observation in the pressurizer at quasi-steady-state, and measurement of pressure, temperature and void fraction at certain fixed locations during both quasi-steady-state and transient experiments are obtained. The IDRIFF code is then applied to supplement the empirical experiment in generating a complete data base, so that extensive theoretical and empirical analyses of the pressurizer behaviour can be systematically performed or verified. The technique of applying the IDRIFF code to simulate both the quasi-steady-state and transient experiment is discussed in detail in the paper. The result of the simulation is in good agreement with measurements taken during the experiment. Analysis of both the empirical and numerical data results in: (1) relationships among void fraction, heater power and steam-bleed flow;(2) a pressurizer flow-regime map and (3) constitutive equations for bubble rising flow and droplet drop flow. This strongly suggests that the approach of extrapolating information obtained from empirical experiment by numerical simulation is a useful method in two-phase flow analysis

  11. Steaming ahead

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    An example of the development of geothermal power in Indonesia is described. Wells are being drilled into the Salak volcano on Java, about 60km south of Jakarta. These let out high pressure hot water trapped 1 to 3km below the surface which can be flashed into steam for driving turbines. The hot water field has already produced 110MW of power since 1994 and is currently being expanded to 330MW. Some details of the drilling and civil engineering are given. Since Indonesia sits on the edge of giant tectonic boundary known as the ''Pacific ring of fire'', the potential for further development is enormous. Ultimately volcanic activity could release an estimated 27,000MW capacity. More realistically, 2,000MW of crustal power by 2020 is spoken of. (UK)

  12. Models and criteria for prediction of Deflagration-to-Detonation Transition (DDT) in hydrogen-air-steam systems under severe accident conditions. Final report

    International Nuclear Information System (INIS)

    Klein, R.; Rehm, W.

    1999-01-01

    The European Commission in Brussels supported a joint project on Deflagration-to-Detonation Transition (DDT) studies for hydrogen safety within the framework programme on nuclear fission safety. The project was initiated by the Forschungszentrum Juelich based on the results of a pilot project. The following main project was coordinated by the Freie Universitaet Berlin involving seven european partners. The partners came from universities, research centers and industry, as follows: FU-Berlin, RWTH-Aachen, CNRS-Marseille, IPSN-Saclay, FZ-Juelich, FZ-Karlsruhe, and NNC-Knutsford, which worked closely together. The working period was two years (1997-1998). The aim of the project was to develop models and criteria for prediction of deflagration-to-detonation transition (DDT) in hydrogen-air-steam systems under severe accident conditions. The results obtained are documented in this final report, which was finished in 1999. The report consists of seven chapters, concerning: - Introduction - Experimental Investigations - Modelling and Numerics - Validation - Mitigation - Further Deliverables - Summary and Conclusion. The final report presents special experimental, theoretical, and computational aspects of the complex DDT phenomena for hydrogen safety studies, and it should be a solid basis for end user applications and further developments. (orig.)

  13. Downhole Upgrading of Orinoco Basin Extra-Heavy Crude Oil Using Hydrogen Donors under Steam Injection Conditions. Effect of the Presence of Iron Nanocatalysts

    Directory of Open Access Journals (Sweden)

    Cesar Ovalles

    2015-03-01

    Full Text Available An extra-heavy crude oil underground upgrading concept and laboratory experiments are presented which involve the addition of a hydrogen donor (tetralin to an Orinoco Basin extra-heavy crude oil under steam injection conditions (280–315 °C and residence times of at least 24-h. Three iron-containing nanocatalysts (20 nm, 60 nm and 90 nm were used and the results showed increases of up to 8° in API gravity, 26% desulfurization and 27% reduction in the asphaltene content of the upgraded product in comparison to the control reaction using inert sand. The iron nanocatalysts were characterized by SEM, XPS, EDAX, and Mössbauer spectroscopy before and after the upgrading reactions. The results indicated the presence of hematite (Fe2O3 as the predominant iron phase. The data showed that the catalysts were deactivating by particle sintering (~20% increase in particle size and also by carbon deposition. Probable mechanisms of reactions are proposed.

  14. Two-phase flow structure in large diameter pipes

    International Nuclear Information System (INIS)

    Smith, T.R.; Schlegel, J.P.; Hibiki, T.; Ishii, M.

    2012-01-01

    Highlights: ► Local profiles of various quantities measured in large diameter pipe. ► Database for interfacial area in large pipes extended to churn-turbulent flow. ► Flow regime map confirms previous models for flow regime transitions. ► Data will be useful in developing interfacial area transport models for large pipes. - Abstract: Flow in large pipes is important in a wide variety of applications. In the nuclear industry in particular, understanding of flow in large diameter pipes is essential in predicting the behavior of reactor systems. This is especially true of natural circulation Boiling Water Reactor (BWR) designs, where a large-diameter chimney above the core provides the gravity head to drive circulation of the coolant through the reactor. The behavior of such reactors during transients and during normal operation will be predicted using advanced thermal–hydraulics analysis codes utilizing the two-fluid model. Essential to accurate two-fluid model calculations is reliable and accurate computation of the interfacial transfer terms. These interfacial transfer terms can be expressed as the product of one term describing the potential driving the transfer and a second term describing the available surface area for transfer, or interfacial area concentration. Currently, the interfacial area is predicted using flow regime dependent empirical correlations; however the interfacial area concentration is best computed through the use of the one-dimensional interfacial area transport equation (IATE). To facilitate the development of IATE source and sink term models in large-diameter pipes a fundamental understanding of the structure of the two-phase flow is essential. This understanding is improved through measurement of the local void fraction, interfacial area concentration and gas velocity profiles in pipes with diameters of 0.102 m and 0.152 m under a wide variety of flow conditions. Additionally, flow regime identification has been performed to

  15. The testing of a steam-water separating device used for vertical steam generators

    International Nuclear Information System (INIS)

    Ding Xunshen; Cui Baoyuan; Xue Yunkui; Liu Shixun

    1989-01-01

    The air-water screening tests of a steam-water separating device used for vertical steam generators at low pressure are introduced. The article puts emphasis on the qualification test of the steam-water separating device at hot conditions in a high temperature and pressure water test rig. The performance of the comprehensive test of the steam-water separating device indicates that the humidity of the steam at the drier exit is much less than the specified amount of 0.25%

  16. Microgravity two-phase flow and heat transfer

    CERN Document Server

    Gabriel, Kamiel S

    2007-01-01

    Advances in understanding the behaviour of multiphase thermal systems could lead to higher efficiency energy production systems, but such advances have been greatly hindered by the strong effect of gravitational acceleration on the flow. This book presents a coverage of various aspects of two-phase flow behaviour in the virtual absence of gravity.

  17. Approximate characteristics for one-dimensional two-phase flows

    International Nuclear Information System (INIS)

    Sarayloo, A.; Peddleson, J.

    1985-01-01

    An approximate method for determining the characteristics associated with one-dimensional particulate two-phase flow models is presented. The method is based on iteration and is valid for small particulate volume fractions. The method is applied to several special cases involving incompressible particles suspended in a gas. The influences of certain changes in the physical model are investigated

  18. TWO-PHASE EJECTOR of CARBON DIOXIDE HEAT PUMP CALCULUS

    Directory of Open Access Journals (Sweden)

    Sit B.M.

    2010-12-01

    Full Text Available It is presented the calculus of the two-phase ejector for carbon dioxide heat pump. The method of calculus is based on the method elaborated by S.M. Kandil, W.E. Lear, S.A. Sherif, and is modified taking into account entrainment ratio as the input for the calculus.

  19. Stability of equilibria for a two-phase osmosis model

    NARCIS (Netherlands)

    Lippoth, F.; Prokert, G.

    2012-01-01

    For a two-phase moving boundary problem modelling the motion of a semipermeable membrane by osmotic pressure and surface tension, we prove that the manifold of equilibria is locally exponentially attractive. Our method relies on maximal regularity results for parabolic systems with relaxation type

  20. Determination of bubble parameters in two-phase flow

    International Nuclear Information System (INIS)

    Oliveira Lira, C.A.B. de.

    1980-01-01

    A development of a probe-detector system for measurement of bubble parameters like size, rise velocity and void fraction in two-phase flow is presented. The method uses an electro resistivity probe and a compact electronic circuit has been developed for obtain this purpose. (author)