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Sample records for heated rod bundle

  1. Single-phase convective heat transfer in rod bundles

    International Nuclear Information System (INIS)

    Holloway, Mary V.; Beasley, Donald E.; Conner, Michael E.

    2008-01-01

    The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids

  2. Single-phase convective heat transfer in rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Holloway, Mary V. [Mechanical Engineering Department, United States Naval Academy, 590 Holloway Rd., Annapolis, MD 21402 (United States)], E-mail: holloway@usna.edu; Beasley, Donald E. [Mechanical Engineering Department, Clemson University, Clemson, SC 29634 (United States); Conner, Michael E. [Westinghouse Nuclear Fuel, 5801 Bluff Road, Columbia, SC 29250 (United States)

    2008-04-15

    The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids.

  3. FLECHT-SEASET 21-rod bundle flow blockage heat transfer during reflood

    International Nuclear Information System (INIS)

    Loftus, M.; Hochreiter, L.; Lee, N.

    1983-01-01

    The effect of various flow blockage shapes and distributions during a PWR reflood was investigated using six 21-rod bundles with full length, internally heated, cosine power-shaped electrical rods. The flow blockage shapes, simulating the fuel rod clad ballooning, were made of thin-wall stainless steel tubes hydroformed into a short, concentric shape and along, nonconcentric shape. The blockage sleeves were distributed both coplanar, with all sleeves located at the same elevation, and non-coplanar. The initial and boundary conditions were varied to include parametric effects of pressure, inlet water temperature, and primarily, flooding rate. The initial mid-plane rod temperature was 871 0 C (1600 0 F) in all tests. Rod and vapor temperature measurements were made throughout the rod bundle with emphasis on the blockage region. The rod heat transfer downstream of the blockage was found to be greater for rods in a blocked bundle than for similar rods in an unblocked bundle. The heat transfer improvement decreases both with time after flood initiation and as the distance increased downstream of the blockage. The improvement in the heat transfer is attributed primarily to the breakup of the water droplets entrained in the steam flow. The smaller droplets subsequently evaporate and desuperheat the steam, which then improves the heat transfer between the rods and the steam in and downstream of the blockage zone

  4. Heat transfer coefficient testing in nuclear fuel rod bundles with mixing vane grids

    International Nuclear Information System (INIS)

    Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

    2005-01-01

    An air heat transfer test facility was developed to test the heat transfer downstream of support grids in simulated PWR nuclear fuel rod bundles. The goal of this testing is to study the single-phase heat transfer coefficients downstream of grids with mixing vanes in a square-pitch rod bundle. The technique developed utilizes fully-heated grid spans and a specially designed thermocouple holder that can be moved axially down the rod bundle and aximuthally within a test rod. From this testing, the axial and aximuthally varying heat transfer coefficient can be determined. Different grid designs are tested and compared to determine the heat transfer enhancement associated with key grid features such as mixing vanes. (author)

  5. Experimental study on the effect of heat flux tilt on rod bundle dryout limitation

    Energy Technology Data Exchange (ETDEWEB)

    Sugawara, S; Terunuma, K; Kamoshida, H [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-12-31

    The effect of heat flux tilt on rod bundle dryout limitation was studied experimentally using a full-scale mock-up test facility and simulated 36-rod fuel bundles in which heater pins have azimuthal nonuniform heat flux distribution (i.e., heat flux tilt). Experimental results for typical lateral power distribution in the bundle indicate that the bundle dryout power with azimuthal heat flux tilt is higher than that without azimuthal heat flux tilt in the entire experimental range. Consequently, it is concluded that the dryout experiment using the test bundle with heater pins which has circumferentially uniform heat flux distribution gives conservative results for the usual lateral power distribution in a bundle in which the relative power of outermost-circle fuel rods is higher than those of middle- and inner-circle ones. (author). 15 refs., 2 tabs., 8 figs.

  6. Experimental study on the effect of heat flux tilt on rod bundle dryout limitation

    International Nuclear Information System (INIS)

    Sugawara, S.; Terunuma, K.; Kamoshida, H.

    1995-01-01

    The effect of heat flux tilt on rod bundle dryout limitation was studied experimentally using a full-scale mock-up test facility and simulated 36-rod fuel bundles in which heater pins have azimuthal nonuniform heat flux distribution (i.e., heat flux tilt). Experimental results for typical lateral power distribution in the bundle indicate that the bundle dryout power with azimuthal heat flux tilt is higher than that without azimuthal heat flux tilt in the entire experimental range. Consequently, it is concluded that the dryout experiment using the test bundle with heater pins which has circumferentially uniform heat flux distribution gives conservative results for the usual lateral power distribution in a bundle in which the relative power of outermost-circle fuel rods is higher than those of middle- and inner-circle ones. (author). 15 refs., 2 tabs., 8 figs

  7. Experimental study of nonequilibrium post-chf heat transfer in rod bundles

    International Nuclear Information System (INIS)

    Unal, C.; Tuzla, K.; Badr, O.; Neti, S.; Chen, J.

    1986-01-01

    Verifications and improvements of nonequilibrium heat transfer models, for post-critical-heat-flux convective boiling, has been greatly affected by the lack of experimental data regarding the degree of thermodynamic nonequilibrium. Recent studies had been successful in measuring vapor superheats in a vertical single tube. This paper extends the nonequilibrium convective boiling data to a rod bundle geometry. Vapor superheat measurements were obtained in a rod bundle with nine heated rods and a heated shroud. Tests were carried out with water at low mass fluxes with a wide range of dryout conditions. Significant nonequilibrium was observed, with vapor superheats of up to 600 0 C. Parametric effects of mass flux, heat flux and inlet conditions on vapor superheat are presented

  8. Rod Bundle Heat Transfer: Steady-State Steam Cooling Experiments

    International Nuclear Information System (INIS)

    Spring, J.P.; McLaughlin, D.M.

    2006-01-01

    Through the joint efforts of the Pennsylvania State University and the United States Nuclear Regulatory Commission, an experimental rod bundle heat transfer (RBHT) facility was designed and built. The rod bundle consists of a 7 x 7 square pitch array with spacer grids and geometry similar to that found in a modern pressurized water reactor. From this facility, a series of steady-state steam cooling experiments were performed. The bundle inlet Reynolds number was varied from 1 400 to 30 000 over a pressure range from 1.36 to 4 bars (20 to 60 psia). The bundle inlet steam temperature was controlled to be at saturation for the specified pressure and the fluid exit temperature exceeded 550 deg. C in the highest power tests. One important quantity of interest is the local convective heat transfer coefficient defined in terms of the local bulk mean temperature of the flow, local wall temperature, and heat flux. Steam temperatures were measured at the center of selected subchannels along the length of the bundle by traversing miniaturized thermocouples. Using an analogy between momentum and energy transport, a method was developed for relating the local subchannel centerline temperature measurement to the local bulk mean temperature. Wall temperatures were measured using internal thermocouples strategically placed along the length of each rod and the local wall heat flux was obtained from an inverse conduction program. The local heat transfer coefficient was calculated from the data at each rod thermocouple location. The local heat transfer coefficients calculated for locations where the flow was fully developed were compared against several published correlations. The Weisman and El-Genk correlations were found to agree best with the RBHT steam cooling data, especially over the range of turbulent Reynolds numbers. The effect of spacer grids on the heat transfer enhancement was also determined from instrumentation placed downstream of the spacer grid locations. The local

  9. Tabular method of critical heat flux description in square packing rod bundles

    International Nuclear Information System (INIS)

    Bobkov, V.P.; Smogalev, I.P.

    2003-01-01

    Elaborations of harnessing tabular method for the description and calculation of critical heat fluxes in square packing rod bundles are presented. The tabular method for fuel rod triangular assemblies derived from using basic table for critical heat fluxes in triangular fuel assemblies demonstrates good results. For the harnessing tabular method in square packing rod bundles correction functions reflecting specific geometry were found. Comparative evaluations of calculated values for the critical heat fluxes with experimental ones are presented. Good agreement of calculations with experiments is noted in all range of parameters [ru

  10. Effective thermal conductivity of a heat generating rod bundle dissipating heat by natural convection and radiation

    International Nuclear Information System (INIS)

    Senve, Vinay; Narasimham, G.S.V.L.

    2011-01-01

    Highlights: → Transport processes in isothermal hexagonal sheath with 19 heat generating rods is studied. → Correlation is given to predict the maximum temperature considering all transport processes. → Effective thermal conductivity of rod bundle can be obtained using max temperature. → Data on the critical Rayleigh numbers for p/d ratios of 1.1-2.0 is presented. → Radiative heat transfer contributes to heat dissipation of 38-65% of total heat. - Abstract: A numerical study of conjugate natural convection and surface radiation in a horizontal hexagonal sheath housing 19 solid heat generating rods with cladding and argon as the fill gas, is performed. The natural convection in the sheath is driven by the volumetric heat generation in the solid rods. The problem is solved using the FLUENT CFD code. A correlation is obtained to predict the maximum temperature in the rod bundle for different pitch-to-diameter ratios and heat generating rates. The effective thermal conductivity is related to the heat generation rate, maximum temperature and the sheath temperature. Results are presented for the dimensionless maximum temperature, Rayleigh number and the contribution of radiation with changing emissivity, total wattage and the pitch-to-diameter ratio. In the simulation of a larger system that contains a rod bundle, the effective thermal conductivity facilitates simplified modelling of the rod bundle by treating it as a solid of effective thermal conductivity. The parametric studies revealed that the contribution of radiation can be 38-65% of the total heat generation, for the parameter ranges chosen. Data for critical Rayleigh number above which natural convection comes into effect is also presented.

  11. Effect of local heat flux spikes on DNB in non-uniformly heated rod bundles

    International Nuclear Information System (INIS)

    Cadek, F.F.; Hill, K.W.; Motley, F.E.

    1975-02-01

    High pressure water tests were carried out to measure the DNB heat flux using an electrically heated rod bundle in which three adjacent rods had 20 percent heat flux spikes at the axial location where DNB is most likely to occur. This test series was run at the same conditions as those of two earlier test series which had unspiked rods, so that spiked and unspiked runs could be paired and spike effects could thus be isolated. Results are described. 7 references. (U.S.)

  12. Application of tube critical heat flux tables to annuli and rod bundles

    International Nuclear Information System (INIS)

    Ulrych, G.

    1985-01-01

    The purpose of this paper is to show that tables for the critical heat flux (CHF) in tubes have a much wider range of applicability than only to tubes. With the proper choice of a characteristic length replacing the tube diameter as a parameter the validity of the tables can be expanded to more complex geometries. The paper describes how the tables must be applied to annuli or rod bundles. The data base for comparisons is mainly taken from the open literature. For rod bundles the proposed methodology was checked for very different geometries including rod bundles from very tight hexagonal to extremely open square bundle arrays. It is concluded that the tables give reasonable results for a wide range of hydraulic diameters

  13. Heat Transfer Coefficient Variations in Nuclear Fuel Rod Bundles

    International Nuclear Information System (INIS)

    Conner, Michael E.; Holloway, Mary V.

    2007-01-01

    The single-phase heat transfer performance of a PWR nuclear fuel rod bundle is enhanced by the use of mixing vanes attached to the downstream edges of the support grid straps. This improved single-phase performance will delay the onset of nucleate boiling, thereby reducing corrosion and delaying crud-related issues. This paper presents the variation in measured single-phase heat transfer coefficients (HTC) for several grid designs. Then, this variation is compared with observations of actual in-core crud patterns. While crud deposition is a function of a number of parameters including rod heat flux, the HTC is assumed to be a primary factor in explaining why crud deposition is a local phenomenon on nuclear fuel rods. The data from this study will be used to examine this assumption by providing a comparison between HTC variations and crud deposition patterns. (authors)

  14. Local heat transfer coefficient for turbulent flow in rod bundles

    International Nuclear Information System (INIS)

    Fernandez y Fernandez, E.; Carajilescov, P.

    1983-03-01

    The correlation of the local heat transfer coefficients in heated triangular array of rod bundles, in terms of the flow hydrodynamic parameters is presented. The analysis is made first for fluid with Prandtl numbers varying from moderated to high (Pr>0.2), and then extended to fluids with low Prandtl numbers (0.004 [pt

  15. Experimental studies on heat transfer to supercritical water in 2 × 2 rod bundle with two channels

    International Nuclear Information System (INIS)

    Gu, H.Y.; Hu, Z.X.; Liu, D.; Xiao, Y.; Cheng, X.

    2015-01-01

    Highlights: • Heat transfer to supercritical water in a 2 × 2 rod bundle is investigated. • Effects of system parameters on heat transfer in bundle are analyzed. • The test data were compared with twenty heat transfer correlations. - Abstract: The experiment of heat transfer to supercritical water in 2 × 2 rod bundle is performed at Shanghai Jiao Tong University. The test section consists of two channels separated by a square steel assembly box with rounded corners. Water flows downward in the first channel and then turns upward in the second channel to cool the 2 × 2 rod bundle installed inside the assembly box. The bundle consists of four heated rods of 10 mm in O.D. and 1.18 in pitch-to-diameter ratio. The fluid enthalpy in the first channel increases due to the heat transfer through the assembly box when flowing downward. The minimum fluid enthalpy increase in the first channel appears at the pseudo-critical region due to the small temperature difference between the two channels. Effects of various parameters on heat transfer behavior inside the 2 × 2 rod bundle are similar to those observed in tube or annuli. No special phenomenon in heat transfer is observed during the mass flux and power transient. The steady-state heat transfer correlation is applicable to predict the heat transfer in the mass or power transient sequence. In addition, the importance of several dimensionless numbers and the accuracy of 20 heat transfer correlations are assessed. It is concluded that the buoyancy parameter proposed by Cheng et al. (2009) shows unique effect on heat transfer coefficient. Among the 20 selected heat transfer correlations, the correlations of Jackson and Fewster (1975) and Bishop et al. (1964) give the best predictions when compared with the experimental data

  16. Design concept and testing of an in-bundle gamma densitometer for subchannel void fraction measurements in the THTF electrically heated rod bundle

    International Nuclear Information System (INIS)

    Felde, D.K.

    1982-04-01

    A design concept is presented for an in-bundle gamma densitometer system for measurement of subchannel average fluid density and void fraction in rod or tube bundles. This report describes (1) the application of the design concept to the Thermal-Hydraulic Test Facility (THTF) electrically heated rod bundle; and (2) results from tests conducted in the THTF

  17. Experimental investigations of single and two-phase flow in a heated rod bundle

    International Nuclear Information System (INIS)

    Barthel, Frank; Franz, Ronald; Hampel, Uwe; Technische Univ. Dresden

    2013-01-01

    An experimental facility for the study of boiling flows in a 3 x 3 rod bundle geometry was setup. The bundle resembles in essential geometrical parts the geometry in a pressurized water reactor fuel element. The facility is operated with a refrigerant fluid. Beside standard instrumentation for temperature, pressure and flow rate we employed particle image velocimetry for single phase flow studies, gamma ray densitometry for integral gas fraction measurement sand ultrafast X-ray tomography for the study of the void dynamics in the cross-section. Moreover extensive thermo-instrumentation allows axial rod surface temperature measurements for the central heated rod. First results will be discussed in this article. (orig.)

  18. Heat transfer in a seven-rod test bundle with supercritical pressure water (1). Experiments

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Seki, Yohji; Dairaku, Masayuki; Suzuki, Satoshi; Enoeda, Mikio; Akiba, Masato; Mori, H.; Oka, Y.

    2009-01-01

    Heat transfer experiments in a seven-rod test bundle with supercritical pressure water has been carried out. The pressure drop and heat transfer coefficients (HTCs) in the test section are evaluated. In the present limited conditions, difference between HTCs at the surface facing the sub-channel center and those at the surface in the narrowest region between rods is not observed. (author)

  19. Numerical investigation of supercritical water-cooled nuclear reactor in horizontal rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Shang Zhi, E-mail: shangzhi@tsinghua.org.c [Faculty of Engineering, Kingston University, London SW15 3DW (United Kingdom); Science and Technology Facilities Council, Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Lo, Simon, E-mail: simon.lo@uk.cd-adapco.co [CD-adapco, Trident House, Basil Hill Road, Didcot OX11 7HJ (United Kingdom)

    2010-04-15

    The commercial CFD code STAR-CD v4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round rods and rod bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal rods and rod bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. In the rod bundle simulations, it is found that the geometry and orientation of the rod bundle have strong effects on the wall temperature distributions and heat transfers. In one orientation the square bundle has a higher wall temperature difference than other bundles. However, when the bundles are rotated by 90 deg. the highest wall temperature difference is found in the hexagon bundle. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR.

  20. Heat-transfer in a partially-blocked sodium-cooled rod bundle

    International Nuclear Information System (INIS)

    Han, J.T.

    1979-01-01

    Heat transfer coefficients were experimentally determined for 31-rod sodium-cooled bundle with a 6-subchannel central blockage. The Nusselt number is presented as a function of the Peclet number for both the free flow region undisturbed by the blockage and the wake region immediately downstream of the blockage. Results are compared with the existing correlations for liquid metals. The heat transfer coefficient was generally higher in the unblocked free flow region than in the wake region. A leak at the blockage improved the heat transfer coefficient in the wake region

  1. Void fraction distribution in a heated rod bundle under flow stagnation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herrero, V.A.; Guido-Lavalle, G.; Clausse, A. [Centro Atomico Bariloche and Instituto Balseiro, Bariloche (Argentina)

    1995-09-01

    An experimental study was performed to determine the axial void fraction distribution along a heated rod bundle under flow stagnation conditions. The development of the flow pattern was investigated for different heat flow rates. It was found that in general the void fraction is overestimated by the Zuber & Findlay model while the Chexal-Lellouche correlation produces a better prediction.

  2. Characteristics of turbulent velocity and temperature in a wall channel of a heated rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Krauss, T.; Meyer, L. [Forschungszentrum Karlsruhe (Germany)

    1995-09-01

    Turbulent air flow in a wall sub-channel of a heated 37-rod bundle (P/D = 1.12, W/D = 1.06) was investigated. measurements were performed with hot-wire probe with X-wires and a temperature wire. The mean velocity, the mean fluid temperature, the wall shear stress and wall temperature, the turbulent quantities such as the turbulent kinetic energy, the Reynolds-stresses and the turbulent heat fluxes were measured and are discussed with respect to data from isothermal flow in a wall channel and heated flow in a central channel of the same rod bundle. Also, data on the power spectral densities of the velocity and temperature fluctuations are presented. These data show the existence of large scale periodic fluctuations are responsible for the high intersubchannel heat and momentum exchange.

  3. Heat transfer in rod bundles with severe clad deformations

    International Nuclear Information System (INIS)

    Ihle, P.

    1984-04-01

    The content of the paper is focused on heat transfer conditions during the reflood phase of a LOCA in slightly to severely deformed PWR fuel rod bundle geometries. The status of analytical and, especially, of experimental work is described as far as it is possible within this frame. Emphasis is placed on the presentation of the results of ''Flooding Experiments with Blocked Arrays'' (FEBA), a program performed at the Kernforschungszentrum Karlsruhe in the frame of the Project Nuclear Safety (PNS). (orig./WL) [de

  4. CHF prediction in rod bundles using round tube data

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Wallen F.; Veloso, Maria A.F.; Pereira, Cláubia; Costa, Antonella L., E-mail: wallenfds@yahoo.com.br, E-mail: mdora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

    2017-07-01

    The present work concerns the use of 1995 CHF table for uniformly heated round tubes, developed jointly by Canadian and Russian researchers, for the prediction of critical heat fluxes in rod bundles geometries. Comparisons between measured and calculated critical heat fluxes indicate that this table could be applied to rod bundles provided that a suitable correction factor is employed. The tolerance limits associated with the departure from nucleate boiling ratio (DNBR) are evaluated by using statistical analysis. (author)

  5. An overview on rod-bundle thermal-hydraulic analyses

    International Nuclear Information System (INIS)

    Sha, W.T.

    1980-01-01

    Three methods used in rod-bundle thermal-hydraulic analysis are summarized. These methods are: (1) subchannel analysis, (2) porous medium formulation with volume porosity, surface permeability, distributed resistance and distributed heat source (sink) and, (3) bench-mark rod-bundle thermal-hydraulic analysis using a boundary-fitted coordinate system. Basic limitations and merits of each method are delineated. (orig.)

  6. Annular burnout data from rod-bundle experiments

    International Nuclear Information System (INIS)

    Yoder, G.L.; Morris, D.G.; Mullins, C.B.

    1983-01-01

    Burnout data for annular flow in a rod bundle are presented for both transient and steady-state conditions. Tests were performed at the Oak Ridge National Laboratory in the Thermal Hydraulic Test Facility (THTF), a pressurized-water loop containing an electrically heated 64-rod bundle. The bundle configuration is typical of later generation pressurized-water reactors with 17 x 17 fuel arrays. Both axial and radial power profiles are flat. All experiments were carried out in upflow with subcooled inlet conditions, insuring accurate flow measurement. Conditions within the bundle were typical of those which could be encountered during a nuclear reactor loss-of-coolant accident

  7. Flowing and heat transfer characteristics of turbulent flow in typical rod bundles at rolling motion

    International Nuclear Information System (INIS)

    Yan Binghuo; Yu Lei; Gu Hanyang

    2011-01-01

    The influence mechanism of rolling motion on the flowing and heat transfer characteristics of turbulent flow in typical four rod bundles was investigated with Fluent code. The flowing and heat transfer characteristics of turbulent flow in rod bundles can be affected by rolling motion. But the flowing similarity of turbulent flow in adiabatic and non-adiabatic can not be affected. If the rolling period is small, the radial additional force can make the parameter profiles, the turbulent flowing and heat transfer change greatly. At rolling motion, as the pitch to diameter ratio decreases, especially if it is less than 1.1, the flowing and heat transfer of turbulent flow at rolling motion change significantly. The variation of pitch to diameter ratio can change the profiles of secondary flow and turbulent kinetic energy in cross-section greatly. (authors)

  8. Calculation study of nonequilibrium post-CHF heat transfer in rod bundle test using modified RELAP5/MOD2

    International Nuclear Information System (INIS)

    Hassan, Y.A.

    1987-01-01

    To date there is only very limited data for non-equilibrium convective film boiling in rod bundle geometries. A recent nine (3 x 3) rod bundle post-critical-flux (CHF) test from the Lehigh University test facility was simulated using RELAP5/MOD2, to assess its capabilities in predicting the overall convective mechanisms in post-CHF heat transfer in rod bundle geometries. The code calculations were compared with experimental data. The code predicted low vapor superheats and void fraction oscillations. A new interfacial heat transfer between the droplet/steam resulted in a reasonable prediction of vapor superheats. A revised dispersed flow film boiling correlation which accounts for the enhancement of steam convective cooling by droplet-induced turbulence was incorporated in the code. Comparison with the data showed a fair agreement

  9. New models of droplet deposition and entrainment for prediction of CHF in cylindrical rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Haibin, E-mail: hb-zhang@xjtu.edu.cn [School of Chemical Engineering and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom); Hewitt, G.F. [Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom)

    2016-08-15

    Highlights: • New models of droplet deposition and entrainment in rod bundles is developed. • A new phenomenological model to predict the CHF in rod bundles is described. • The present model is well able to predict CHF in rod bundles. - Abstract: In this paper, we present a new set of model of droplet deposition and entrainment in cylindrical rod bundles based on the previously proposed model for annuli (effectively a “one-rod” bundle) (2016a). These models make it possible to evaluate the differences of the rates of droplet deposition and entrainment for the respective rods and for the outer tube by taking into account the geometrical characteristics of the rod bundles. Using these models, a phenomenological model to predict the CHF (critical heat flux) for upward annular flow in vertical rod bundles is described. The performance of the model is tested against the experimental data of Becker et al. (1964) for CHF in 3-rod and 7-rod bundles. These data include tests in which only the rods were heated and data for simultaneous uniform and non-uniform heating of the rods and the outer tube. It was shown that the predicted CHFs by the present model agree well with the experimental data and with the experimental observation that dryout occurred first on the outer rods in 7-rod bundles. It is expected that the methodology used will be generally applicable in the prediction of CHF in rod bundles.

  10. The droplet injection system used in the rod bundle heat transfer facility

    International Nuclear Information System (INIS)

    Frepoli, C.; Andrew, A.J.; Hochreiter, L.E.; Cheung, F.B.

    2001-01-01

    The full text follows. The US Nuclear Regulatory Commission (NRC) and the Pennsylvania State University are currently funding a research program entitled ''Rod Bundle Heat Transfer'' (RBHT). The main objective of the program is to investigate heat transfer during the core reflood period of a hypothetical Large Break Loss of Coolant Accident in a typical nuclear power plant. The RBHT test facility consists of a full-length 7 x 7 rod bundle. Information gathered by the RBHT test facility will be used for improvement of the reflood heat transfer models in the NRC's thermal hydraulic codes. In particular the RBHT data will be used to improve the understanding of individual heat transfer effects to the total rod heat transfer such that compensating errors present in current Best Estimate codes can be significantly reduced. The strategy in developing the test matrix is to use a ''building block'' approach in which simpler experiments are performed first to quantify a particular heat transfer mechanism alone and then the additional complications of the full two-phase flow, reflood film boiling behavior of the test facility are added in later experiments. One of these ''simpler'' experiments will be the injection of known size and velocity liquid droplets into the main stream of superheated steam. The droplet injection system consists of small diameter tubes inserted across the bundle at a given elevation. A number of equal size holes are drilled perpendicular to the surface in a triangular pitch. Water is forced into opposite ends of the tube and ejected from the holes. The injection system was tested using a digital imaging system known as VisiSizer. This system is capable of determining the diameter and velocity of small water droplets using a laser-illuminated digital camera system (LIDCS). Imaging software analyzes the digital images in real time to determine the distributions of droplet size and velocity. Pre-test analysis using COBRA-TF have been conducted to

  11. Subchannel analysis and correlation of the Rod Bundle Heat Transfer (RBHT) steam cooling experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Riley, M.P.; Mohanta, L.; Miller, D.J.; Cheung, F.B. [Pennsylvania State Univ., University Park, PA (United States); Bajorek, S.M.; Tien, K.; Hoxie, C.L. [U.S. Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

    2016-07-15

    A subchannel analysis of the steam cooling data obtained in the Rod Bundle Heat Transfer (RBHT) test facility has been performed in this study to capture the effect of spacer grids on heat transfer. The RBHT test facility has a 7 x 7 rod bundle with heater rods and with seven spacer grids equally spaced along the length of the rods. A method based on the concept of momentum and heat transport analogy has been developed for calculating the subchannel bulk mean temperature from the measured steam temperatures. Over the range of inlet Reynolds number, the local Nusselt number was found to exhibit a minimum value between the upstream and downstream spacer grids. The presence of a spacer grid not only affects the local Nusselt number downstream of the grid but also affects the local Nusselt number upstream of the next grid. A new correlation capturing the effect of Reynolds number on the local flow restructuring downstream as well as upstream of the spacer grids was proposed for the minimum Nusselt number. In addition, a new enhancement factor accounting for the effects of the upstream as well as downstream spacer grids was developed from the RBHT data. The new enhancement factor was found to compare well with the data from the ACHILLLES test facility.

  12. A burnout correlation for flow of boiling water in vertical rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M

    1967-04-15

    The rod bundle burnout correlation described in the present report is a development from our earlier published rod bundle correlation for low pressures. The correlation is based on the Becker round duct correlation and is written on the form x{sub BO} = 0.68*{eta}*{eta}{sub L}*X{sub RD} where x{sub RD} is the burnout steam quality in a round duc at corresponding flow conditions, {eta} is the ratio of heated to total perimeter and {eta}{sub l} is a correction factor, which is a function of q/A only. It is demonstrated that this equation combined with the heat balance equation q/A = G/(4L/D{sub H})*({delta}h{sub SUB} + X{sub BO}*H{sub fg}) predicts the burnout heat fluxes for 312 measurements obtained in our laboratory within a scatter of {+-}7. 5 per cent and with an RMS error of 3.8 per cent. The measurements were obtained in the following ranges of variables. Number of rods n 1, 3, 6 and 7; Rod diameter d{sub i} 10.05 - 13.80 mm; Shroud diameter d{sub o} 17. 42 - 71. 0 mm; Rod clearance s 3.7 - 8.8 mm; Heated length L 608 - 4440 mm; Pressure p 20-71 kg/cm{sup 2}, Inlet sub-cooling {delta}t{sub sub} 3 - 240 deg C; Mass velocity G 80-1,500 kg/m{sup 2}; Burnout heat flux q/A 74-314 W/cm{sup 2}; Burnout steam quality x{sub BO} 0. 1 - 0.55. The correlation shows that the burnout conditions in wide ranges of variables are independent of the inlet sub-cooling and the heated length, and that the effects of mass velocity and pressure are the same in rod bundles and in round tubes. It is also demonstrated that the effects of a radial heat flux variation within the rod bundle can be handled by the correlation by modifying the {eta}-value for the bundle. The rod bundle data presented by Janssen and Kervinen, Hench, Obertelli, Matzner, Haslam, Edwards and Obertelli and Hench and Boehm were also analysed in terms of the measured and predicted burnout heat fluxes. These data covered bundles consisting of 3, 4, 6, 7, 9. 19 and 36 rods and it was found that a very good agreement

  13. A burnout correlation for flow of boiling water in vertical rod bundles

    International Nuclear Information System (INIS)

    Becker, Kurt M.

    1967-04-01

    The rod bundle burnout correlation described in the present report is a development from our earlier published rod bundle correlation for low pressures. The correlation is based on the Becker round duct correlation and is written on the form x BO 0.68*η*η L *X RD where x RD is the burnout steam quality in a round duc at corresponding flow conditions, η is the ratio of heated to total perimeter and η l is a correction factor, which is a function of q/A only. It is demonstrated that this equation combined with the heat balance equation q/A = G/(4L/D H )*(Δh SUB + X BO *H fg ) predicts the burnout heat fluxes for 312 measurements obtained in our laboratory within a scatter of ±7. 5 per cent and with an RMS error of 3.8 per cent. The measurements were obtained in the following ranges of variables. Number of rods n 1, 3, 6 and 7; Rod diameter d i 10.05 - 13.80 mm; Shroud diameter d o 17. 42 - 71. 0 mm; Rod clearance s 3.7 - 8.8 mm; Heated length L 608 - 4440 mm; Pressure p 20-71 kg/cm 2 , Inlet sub-cooling Δt sub 3 - 240 deg C; Mass velocity G 80-1,500 kg/m 2 ; Burnout heat flux q/A 74-314 W/cm 2 ; Burnout steam quality x BO 0. 1 - 0.55. The correlation shows that the burnout conditions in wide ranges of variables are independent of the inlet sub-cooling and the heated length, and that the effects of mass velocity and pressure are the same in rod bundles and in round tubes. It is also demonstrated that the effects of a radial heat flux variation within the rod bundle can be handled by the correlation by modifying the η-value for the bundle. The rod bundle data presented by Janssen and Kervinen, Hench, Obertelli, Matzner, Haslam, Edwards and Obertelli and Hench and Boehm were also analysed in terms of the measured and predicted burnout heat fluxes. These data covered bundles consisting of 3, 4, 6, 7, 9. 19 and 36 rods and it was found that a very good agreement existed between the present correlation and the measurements

  14. Measurements of local temperature distributions in rod bundles with sodium flow

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.; Kolodziej, M.

    1984-12-01

    In an electrically heated 19-rod bundle (P/D = 1.30, W/R = 1.40) with sodium flow the three-dimensional temperature fields in the rod clads were measured. The main characteristics of the test section are three adjacent heater rods in the duct wall zone instrumented on four measuring planes and rotatable by 360 0 under full power conditions; furthermore spacer grids which are axially movable, and a system allowing to bow one heater rod over the last third of its heated length. The results of measurements of the azimuthal temperature variations of the rotatable rods are presented for different operating conditions (80 2 ), different spacer grid positions relative to the measuring planes and different bowing positions of one rod. For better understanding of the experimental results cross sections of the 19-rod bundle were prepared. It became evident, that a well-known bundle geometry is very important for the interpretation of the experimental results. (orig.) [de

  15. Wire-wrapped rod-bundle heat-transfer analysis for LMFBR

    International Nuclear Information System (INIS)

    Wong, C.N.C.; Todreas, N.E.

    1982-07-01

    Helical wire wraps are widely used in the LMFBR fuel and blanket assemblies to provide coolant mixing and maintain proper spacing between fuel pins. The presence of the helical wire, however, may possibly induce heat transfer problems, such as the uncertainty of the maximum clad temperature as a result of the contact between the wires and the pins. In this study, the detailed transient three dimensional velocity and temperature distributions for the coolant around the pin will be determined by solving the governing momentum and energy equation numerically. A computer code HEATRAN has been developed to perform this calculation. Before the computer code HEATRAN is applied to the wire wrapped rod bundle problem, it is used to analyze a wide range of fluid and heat transfer problem to verify its capabilities

  16. New Westinghouse correlation WRB-1 for predicting critical heat flux in rod bundles with mixing vane grids

    International Nuclear Information System (INIS)

    Motley, F.E.; Hill, K.W.; Cadek, F.F.; Shefcheck, J.

    1976-07-01

    A new critical heat flux (CHF) correlation, based on local fluid conditions, has been developed from Westinghouse rod bundle data. This correlation applies to both 0.422 inch and 0.374 inch rod O.D. geometries. It accounts for typical cell and thimble cell effects, uniform and non-uniform heat flux profiles, variations in rod heated length and in grid spacing. The correlation predicts CHF for 1147 data points with a sample mean and standard deviation of measured-to-predicted heat flux ratio of 1.0043 and 0.0873, respectively. It was concluded that to meet the reactor design criterion the minimum DNBR should be 1.17

  17. A critical heat flux approach for square rod bundles using the 1995 Groeneveld CHF table and bundle data of heat transfer research facility

    International Nuclear Information System (INIS)

    Lee, M.

    2000-01-01

    The critical heat flux (CHF) approach using CHF look-up tables has become a widely accepted CHF prediction technique. In these approaches, the CHF tables are developed based mostly on the data bank for flow in circular tubes. A set of correction factors was proposed by Groeneveld et al. [Groeneveld, D.C., Cheng, S.C., Doan, T. (1986)] to extend the application of the CHF table to other flow situations including flow in rod bundles. The proposed correction factors are based on a limited amount of data not specified in the original paper. The CHF approach of Groeneveld and co-workers is extensively used in the thermal hydraulic analysis of nuclear reactors. In 1996, Groeneveld et al. proposed a new CHF table to predict CHF in circular tubes [Groeneveld, D.C., et al., 1996. The 1995 look-up table for Critical Heat Flux. Nucl. Eng. Des. 163(1), 23]. In the present study, a set of correction factors is developed to extend the applicability of the new CHF table to flow in rod bundles of square array. The correction factors are developed by minimizing the statistical parameters of the ratio of the measured and predicted bundle CHF data from the Heat Transfer Research Facility. The proposed correction factors include: the hydraulic diameter factor (K hy ), the bundle factor (K bf ), the heated length factor (K hl ), the grid spacer factor (K sp ), the axial flux distribution factors (K nu ), the cold wall factor (K cw ) and the radial power distribution factor (K rp ). The value of constants in these correction factors is different when the heat balance method (HBM) and direct substitution method (DSM) are adopted to predict the experimental results of HTRF. With the 1995 Groeneveld CHF Table and the proposed correction factors, the average relative error is 0.1 and 0.0% for HBM and DSM, respectively, and the root mean square (RMS) error is 31.7% in DSM and 17.7% in HBM for 9852 square array data points of HTRF. (orig.)

  18. Critical heat flux tests for self-spaced square finned 7 fuel rod bundle

    International Nuclear Information System (INIS)

    Moon, Sang Ki; Chun, Se Young; Choi, Ki Young; Park, Jong Kuk; Hwang, Dae Hyun; Zee, Sung Quun; Kim, Keung Koo

    2001-09-01

    Now, KAERI is developing a new advanced reactor aimed at achieving highly enhanced safety and reliability, and improved economics. SSF (Self-Spaced Square Finned) fuel rod bundle is considered as a suitable one for the new advanced reactor. The SSF fuel rods have rectangular shapes and four fins at the corners, and are arranged in triangular geometry. While the SSF fuel rod bundle is considered to have enhanced cooling efficiency, the correlations used for commercial PWR might be able to be applied. The application results of some conventional correlations show that the SSF fuel rod bundle show an enhanced CHF performance about 10 to 40 %. When some conventional CHF correlations are applied to CHF data with a similar geometry to the SSF fuel rod bundle, conventional CHF correlations including a correlation developed in Russia are judged not to be suitable for the development of SSF fuel rod bundle and for the use in a safety analysis code. From CHF experiments for SSF 7 fuel rod bundle performed in KAERI, the following results are obtained: the CHF increases with increasing mass flux, and the CHF increasing rate decreases at high mass flux conditions. The exit quality decreases with increasing mass flux. The overall effect of the mass flux on the CHF and exit quality coincides with previous understanding. Compared to the CHF data of IPPE with the same system pressure and inlet temperature, the CHF data of KAERI show the similar values. Thus, the reliability of IPPE CHF data can be confirmed indirectly

  19. On the calculation of flow and heat transfer characteristics for CANDU-type 19-rod fuel bundles

    International Nuclear Information System (INIS)

    Yuh-Shan Yueh; Ching-Chang Chieng

    1987-01-01

    A numerical study is reported of flow and heat transfer in a CANDU-type 19 rod fuel bundle. The flow domain of interest includes combinations of trangular, square, and peripheral subchannels. The basic equations of momentum and energy are solved with the standard k--ε model of turbulence. Isotropic turbulent viscosity is assumed and no secondary flow is considered for this steady-state, fully developed flow. Detailed velocity and temperature distributions with wall shear stress and Nusselt number distributions are obtained for turbulent flow of Re = 4.35 x 10 4 , 10 5 , 2 x 10 5 , and for laminar flow of Re--2400. Friction factor and heat transfer ceofficients of various subchannels inside the full bundle are compared with those of infinite rod arrays of triangular or square arrangements. The calculated velocity contours of peripheral subchannel agreed reasonably with measured data

  20. Substantiation and verification of the heat exchange crisis model in a rod bundles by means of the KORSAR thermohydraulic code

    International Nuclear Information System (INIS)

    Bobkov, V.P.; Vinogradov, V.N.; Efanov, A.D.; Sergeev, V.V.; Smogalev, I.P.

    2003-01-01

    The results of verifying the model for calculating the heat exchange crisis in the uniformly heated rod bundles, realized in the calculation code of the improved evaluation KORSAR, are presented. The model for calculating the critical heat fluxes in this code is based on the tabular method. The experimental data bank of the Branch base center of the thermophysical data GNTs RF - FEhI for the rod bundles, structurally similar to the WWER fuel assemblies, was used by the verification within the wide range of parameters: pressure from 0.11 up to 20 MPa and mass velocity from 5- up to 5000 kg/(m 2 s) [ru

  1. Rod bundle burnout data and correlation comparisons

    International Nuclear Information System (INIS)

    Yoder, G.L.; Morris, D.G.; Mullins, C.B.

    1985-01-01

    Rod bundle burnout data from 30 steady-state and 3 transient tests were obtained from experiments performed in the Thermal Hydraulic Test Facility at the Oak Ridge National Laboratory. The tests covered a parameter range relevant to intact core reactor accidents ranging from large break to small break loss-ofcoolant conditions. Instrumentation within the 64-rod test section indicated that burnout occurred over an axial range within the bundle. The distance from the point where the first dry rod was detected to the point where all rods were dry was up to 60 cm in some of the tests. The burnout data should prove useful in developing new correlations for use in reactor thermalhydraulic codes. Evaluation of several existing critical heat flux correlations using the data show that three correlations, the Barnett, Bowring, and Katto correlations, perform similarly and correlate the data better than the Biasi correlation

  2. Heat Transfer Enhancement By Three-Dimensional Surface Roughness Technique In Nuclear Fuel Rod Bundles

    Science.gov (United States)

    Najeeb, Umair

    This thesis experimentally investigates the enhancement of single-phase heat transfer, frictional loss and pressure drop characteristics in a Single Heater Element Loop Tester (SHELT). The heater element simulates a single fuel rod for Pressurized Nuclear reactor. In this experimental investigation, the effect of the outer surface roughness of a simulated nuclear rod bundle was studied. The outer surface of a simulated fuel rod was created with a three-dimensional (Diamond-shaped blocks) surface roughness. The angle of corrugation for each diamond was 45 degrees. The length of each side of a diamond block is 1 mm. The depth of each diamond block was 0.3 mm. The pitch of the pattern was 1.614 mm. The simulated fuel rod had an outside diameter of 9.5 mm and wall thickness of 1.5 mm and was placed in a test-section made of 38.1 mm inner diameter, wall thickness 6.35 mm aluminum pipe. The Simulated fuel rod was made of Nickel 200 and Inconel 625 materials. The fuel rod was connected to 10 KW DC power supply. The Inconel 625 material of the rod with an electrical resistance of 32.3 kO was used to generate heat inside the test-section. The heat energy dissipated from the Inconel tube due to the flow of electrical current flows into the working fluid across the rod at constant heat flux conditions. The DI water was employed as working fluid for this experimental investigation. The temperature and pressure readings for both smooth and rough regions of the fuel rod were recorded and compared later to find enhancement in heat transfer coefficient and increment in the pressure drops. Tests were conducted for Reynold's Numbers ranging from 10e4 to 10e5. Enhancement in heat transfer coefficient at all Re was recorded. The maximum heat transfer co-efficient enhancement recorded was 86% at Re = 4.18e5. It was also observed that the pressure drop and friction factor increased by 14.7% due to the increased surface roughness.

  3. Hydrodynamic behavior of a bare rod bundle

    International Nuclear Information System (INIS)

    Bartzis, J.G.; Todreas, N.E.

    1977-06-01

    The temperature distribution within the rod bundle of a nuclear reactor is of major importance in nuclear reactor design. However temperature information presupposes knowledge of the hydrodynamic behavior of the coolant which is the most difficult part of the problem due to complexity of the turbulence phenomena. In the present work a 2-equation turbulence model--a strong candidate for analyzing actual three dimensional turbulent flows--has been used to predict fully developed flow of infinite bare rod bundle of various aspect ratios (P/D). The model has been modified to take into account anisotropic effects of eddy viscosity. Secondary flow calculations have been also performed although the model seems to be too rough to predict the secondary flow correctly. Heat transfer calculations have been performed to confirm the importance of anisotropic viscosity in temperature predictions. All numerical calculations for flow and heat have been performed by two computer codes based on the TEACH code. Experimental measurements of the distribution of axial velocity, turbulent axial velocity, turbulent kinetic energy and radial Reynolds stresses were performed in the developing and fully developed regions. A 2-channel Laser Doppler Anemometer working on the Reference mode with forward scattering was used to perform the measurements in a simulated interior subchannel of a triangular rod array with P/D = 1.124. Comparisons between the analytical results and the results of this experiment as well as other experimental data in rod bundle array available in literature are presented. The predictions are in good agreement with the results for the high Reynolds numbers

  4. Calculations of combined radiation and convection heat transfer in rod bundles under emergency cooling conditions

    International Nuclear Information System (INIS)

    Sun, K.H.; Gonzalez-Santalo, J.M.; Tien, C.L.

    1976-01-01

    A model has been developed to calculate the heat transfer coefficients from the fuel rods to the steam-droplet mixture typical of Boiling Water Reactors under Emergency Core Cooling System (ECCS) operation conditions during a postulated loss-of-coolant accident. The model includes the heat transfer by convection to the vapor, the radiation from the surfaces to both the water droplets and the vapor, and the effects of droplet evaporation. The combined convection and radiation heat transfer coefficient can be evaluated with respect to the characteristic droplet size. Calculations of the heat transfer coefficient based on the droplet sizes obtained from the existing literature are consistent with those determined empirically from the Full-Length-Emergency-Cooling-Heat-Transfer (FLECHT) program. The present model can also be used to assess the effects of geometrical distortions (or deviations from nominal dimensions) on the heat transfer to the cooling medium in a rod bundle

  5. An experimental investigation of supercritical heat transfer in a three-rod bundle equipped with wire-wrap and grid spacers and cooled by carbon dioxide

    Energy Technology Data Exchange (ETDEWEB)

    Eter, Ahmad, E-mail: eng.eter@yahoo.com; Groeneveld, Dé, E-mail: degroeneveld@gmail.com; Tavoularis, Stavros, E-mail: stavros.tavoularis@uottawa.ca

    2016-07-15

    Highlights: • Heat transfer at supercritical pressures was studied experimentally in a three-rod bundle equipped with wire-wrap spacers or grid spacers. • Heat transfer deterioration occurred near the heated inlet under certain conditions. • Normal heat transfer was generally comparable to that in a tube and the predictions of a correlation. - Abstract: Heat transfer measurements in a three-rod bundle equipped with wire-wrap and grid spacers were obtained at supercritical pressures in the Supercritical University of Ottawa Loop (SCUOL). The tests were performed using carbon dioxide, as a surrogate fluid for water, flowing upwards for wide ranges of conditions, including conditions equivalent to the nominal and near-normal operating conditions of the proposed Canadian Super-Critical Water-Cooled Reactor. The test section contained three heated rods and three unheated rod segments with an outer diameter of 10 mm and a pitch-to-diameter ratio of 1.14; the heated length was 1500 mm. Detailed surface temperature measurements along and around the three heated rods were collected using internally traversed thermocouples. The following ranges of test conditions were covered, with equivalent water conditions given inside parentheses: pressure from 6.6 to 8.36 MPa (19.7–25 MPa); inlet temperature from 11 to 30 °C (330–371 °C); mass flux from 200 to 1175 kg m{sup −2} s{sup −1} (340–1822 kg m{sup −2} s{sup −1}); and wall heat flux from 1 to 175 kW m{sup −2} (11–1847 kW m{sup −2}). For one set of tests, the heated rods were fitted with a 1.3 mm OD wire wrap, having an axial pitch of 200 mm along the entire heated length; for a second set, the heated rods were fitted with grid spacers having a 5.3% flow blockage and located at 500 mm axial intervals. The effects of spacer configuration on heat transfer at supercritical pressures were documented and analyzed. The observed experimental trends were compared to those obtained in a experiment in a heated

  6. Critical heat flux near the critical pressure in heater rod bundle cooled by R-134A fluid: Effects of unheated rods and spacer grid

    International Nuclear Information System (INIS)

    Chun, Se-Y.; Shin, C.W.; Hong, S. D.; Moon, S. K.

    2007-01-01

    A supercritical-pressure light water reactor (SCWR) is currently investigated as the next generation nuclear reactors. The SCWR, which is operated above the thermodynamic critical point of water (647 K, 22.1 MPa), have advantages over conventional light water reactors in terms of thermal efficiency as well as in compactness and simplicity. Many experimental studies have been performed on heat transfer in the boiler tubes of supercritical fossil fire power plants (FPPs). However, the thermal-hydraulic conditions of the SCWR core are different from those of the FPP boiler. In the SCWR core, the heat transfer to the cooling water occurs on the outside surface of fuel rods in rod bundle with spacers. In addition, the experimental studies in which the critical heat flux (CHF) has been carefully measured near the critical pressure have never yet been carried out, as far as we know. Therefore, we have recently conducted the CHF experiments with a vertical 5x5 heater rod bundle cooled by R- 134a fluid. The purpose of this work is to find out some novel knowledge for the CHF near the critical pressure, based on more careful experiments. The outer diameter, heated length and rod pitch of the heater rods are 9.5, 2000 and 12.85 mm, respectively. The critical power has been measured in a range of the pressure of 2.474.03 MPa (the critical pressure of R-134a is 4.059 MPa), the mass flux 502000 kg/m 2 s, and the inlet subcooling 4084 kJ/kg. For the mass fluxes of not less than 550 kg/m 2 s, the critical power decreases monotonously up to the pressure of about 3.63.8 MPa with increasing pressure, and then fall sharply at about 3.83.9 MPa as if the values of the critical power converge on zero at the critical pressure. For the low mass fluxes of 50 to 250 kg/m 2 , the sharp decreasing trend of the critical power near the critical pressure is not observed. The CHF phenomenon near the critical pressure no longer leads to an inordinate increase in the heated wall temperature such as

  7. Investigation of Swirling Flow in Rod Bundle Subchannels Using Computational Fluid Dynamics

    International Nuclear Information System (INIS)

    Holloway, Mary V.; Beasley, Donald E.; Conner, Michael E.

    2006-01-01

    The fluid dynamics for turbulent flow through rod bundles representative of those used in pressurized water reactors is examined using computational fluid dynamics (CFD). The rod bundles of the pressurized water reactor examined in this study consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids are often used to create swirling flow in the rod bundle in an effort to improve the heat transfer characteristics for the rod bundle during both normal operating conditions and in accident condition scenarios. Computational fluid dynamics simulations for a two subchannel portion of the rod bundle were used to model the flow downstream of a split-vane pair support grid. A high quality computational mesh was used to investigate the choice of turbulence model appropriate for the complex swirling flow in the rod bundle subchannels. Results document a central swirling flow structure in each of the subchannels downstream of the split-vane pairs. Strong lateral flows along the surface of the rods, as well as impingement regions of lateral flow on the rods are documented. In addition, regions of lateral flow separation and low axial velocity are documented next to the rods. Results of the CFD are compared to experimental particle image velocimetry (PIV) measurements documenting the lateral flow structures downstream of the split-vane pairs. Good agreement is found between the computational simulation and experimental measurements for locations close to the support grid. (authors)

  8. Study on nonstationary convective heat transfer in annular channels and rod bundles in conditions of arbitrary variation of heat duty in time and length

    International Nuclear Information System (INIS)

    Kuznetsov, Yu.N.; Kalinin, E.I.; Naumov, M.A.

    1980-01-01

    The effect of variability of heat duty on the characteristics of heat exchange in ring channels and rod bundles is investigated with analytical methods. The plotting of calculation formulae for non-stationary heat exchange in an annular channel at a jump of heat duty is carried out on the basis of the method of the effect function. The formulae obtained permit to accomplish technical calculations of the processes of non-stationary heat exchange in annular channels in the case of any alterations of thermal duty in time, at any moment of time, for any channel cross section (including the entrance heat section) in a wide range of geometric and regime parameters of the turbulent current of a coolant. According to preliminary estimates, calculation results differ from the results oi a numerical solution less than 5%. The approach considered permits to transfer the data on the non-stationary heat exchange in annular channels in the case of changing the heat duty in time, in the case of a non-stationary heat exchange in longitudinally flown not very dense and infinite rod bundles

  9. BUSH: A computer code for calculating steady state heat transfer in LWR rod bundles under accident conditions

    International Nuclear Information System (INIS)

    Shepherd, I.M.

    1982-01-01

    The computer code BUSH has been developed for the calculation of steady state heat transfer in a rod bundle. For a given power, flow and geometry it can calculate the temperatures in the rods, coolant and shroud assuming that at any axial level each rod can be described by one temperature and the coolant fluid is also radially uniform at this level. Heat transfer by convection and radiation are handled and the geometry is flexible enough to model nearly all types of envisaged shroud design for the SUPERSARA test series. The modular way in which BUSH has been written makes it suitable for future development, either within the present BUSH framework or as part of a more advanced code

  10. Numerical prediction of critical heat flux in nuclear fuel rod bundles with advanced three-fluid multidimensional porous media based model

    International Nuclear Information System (INIS)

    Zoran Stosic; Vladimir Stevanovic

    2005-01-01

    Full text of publication follows: The modern design of nuclear fuel rod bundles for Boiling Water Reactors (BWRs) is characterised with increased number of rods in the bundle, introduced part-length fuel rods and a water channel positioned along the bundle asymmetrically in regard to the centre of the bundle cross section. Such design causes significant spatial differences of volumetric heat flux, steam void fraction distribution, mass flux rate and other thermal-hydraulic parameters important for efficient cooling of nuclear fuel rods during normal steady-state and transient conditions. The prediction of the Critical Heat Flux (CHF) under these complex thermal-hydraulic conditions is of the prime importance for the safe and economic BWR operation. An efficient numerical method for the CHF prediction is developed based on the porous medium concept and multi-fluid two-phase flow models. Fuel rod bundle is observed as a porous medium with a two-phase flow through it. Coolant flow from the bundle entrance to the exit is characterised with the subsequent change of one-phase and several two-phase flow patterns. One fluid (one-phase) model is used for the prediction of liquid heating up in the bundle entrance region. Two-fluid modelling approach is applied to the bubbly and churn-turbulent vapour and liquid flows. Three-fluid modelling approach is applied to the annular flow pattern: liquid film on the rods wall, steam flow and droplets entrained in the steam stream. Every fluid stream in applied multi-fluid models is described with the mass, momentum and energy balance equations. Closure laws for the prediction of interfacial transfer processes are stated with the special emphasis on the prediction of the steam-water interface drag force, through the interface drag coefficient, and droplets entrainment and deposition rates for three-fluid annular flow model. The model implies non-equilibrium thermal and flow conditions. A new mechanistic approach for the CHF prediction

  11. Experimental heat transfer in tube bundle

    International Nuclear Information System (INIS)

    Khattab, M.; Mariy, A.; Habib, M.

    1983-01-01

    Previous work has looked for the problem of heat transfer with flow parallel to rod bundle either by treating each rod individually as a separate channel or by treating the bundle as one unit. The present work will consider the existence of both the central and corner rods simultaneously inside the cluster itself under the same working conditions. The test section is geometrically similar to the fuel assembly of the Egyptian Research Reactor-1. The hydro-thermal performance of bundle having 16 - stainless steel tubes arranged in square array of 1.5 pitch to diameter ratio is investigated. Surface temperature and pressure distributions are determined. Average heat transfer coefficient for both central and corner tubes are correlated. Also, pressure drop and friction factor correlations are predicted. The maximum experimental range of the measured parameters are determined in the nonboiling region at 1400 Reynolds number and 3.64 W/cm 2 . It is found that the average heat transfer coefficient of the central tube is higher than that of the corner tube by 27%. Comparison with the previous work shows satisfactory agreement particularly with the circular tubes correlation - Dittus et al. - at 104 Reynolds number

  12. Hydrodynamic behavior of a bare rod bundle. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Bartzis, J.G.; Todreas, N.E.

    1977-06-01

    The temperature distribution within the rod bundle of a nuclear reactor is of major importance in nuclear reactor design. However temperature information presupposes knowledge of the hydrodynamic behavior of the coolant which is the most difficult part of the problem due to complexity of the turbulence phenomena. In the present work a 2-equation turbulence model--a strong candidate for analyzing actual three dimensional turbulent flows--has been used to predict fully developed flow of infinite bare rod bundle of various aspect ratios (P/D). The model has been modified to take into account anisotropic effects of eddy viscosity. Secondary flow calculations have been also performed although the model seems to be too rough to predict the secondary flow correctly. Heat transfer calculations have been performed to confirm the importance of anisotropic viscosity in temperature predictions. All numerical calculations for flow and heat have been performed by two computer codes based on the TEACH code. Experimental measurements of the distribution of axial velocity, turbulent axial velocity, turbulent kinetic energy and radial Reynolds stresses were performed in the developing and fully developed regions. A 2-channel Laser Doppler Anemometer working on the Reference mode with forward scattering was used to perform the measurements in a simulated interior subchannel of a triangular rod array with P/D = 1.124. Comparisons between the analytical results and the results of this experiment as well as other experimental data in rod bundle array available in literature are presented. The predictions are in good agreement with the results for the high Reynolds numbers.

  13. The Preliminary Study for Numerical Computation of 37 Rod Bundle in CANDU Reactor

    International Nuclear Information System (INIS)

    Jeon, Yu Mi; Park, Joo Hwan

    2010-09-01

    A typical CANDU 6 fuel bundle consists of 37 fuel rods supported by two endplates and separated by spacer pads at various locations. In addition, the bearing pads are brazed to each outer fuel rod with the aim of reducing the contact area between the fuel bundle and the pressure tube. Although the recent progress of CFD methods has provided opportunities for computing the thermal-hydraulic phenomena inside of a fuel channel, it is yet impossible to reflect numerical computations on the detailed shape of rod bundle due to challenges with computing mesh and memory capacity. Hence, the previous studies conducted a numerical computation for smooth channels without considering spacers and bearing pads. But, it is well known that these components are an important factor to predict the pressure drop and heat transfer rate in a channel. In this study, the new computational method is proposed to solve complex geometry such as a fuel rod bundle. Before applying a solution to the problem of the 37 rod bundle, the validity and the accuracy of the method are tested by applying the method to simple geometry. The split channel method has been proposed with the aim of computing the fully shaped CANDU fuel channel with detailed components. The validity was tested by applying the method to the single channel problem. The average temperature have similar values for the considered two methods, while the local temperature shows a slight difference by the effect of conduction heat transfer in the solid region of a rod. Based on the present result, the calculation for the fully shaped 37-rod bundle is scheduled for future work

  14. Effects of fuel relocation on reflood in a partially-blocked rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Jae [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseong-gu, Daejeon 34134 (Korea, Republic of); Kim, Jongrok; Kim, Kihwan; Bae, Sung Won [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Thermal-Hydraulic Safety Research Division, Korea Atomic Energy Research Division, 111 Daedeok-daero, Yuseong-gu, Daejeon 34057 (Korea, Republic of)

    2017-02-15

    Ballooning of the fuel rods has been an important issue, since it can influence the coolability of the rod bundle in a large-break loss-of-coolant accident (LBLOCA). Numerous past studies have investigated the effect of blockage geometry on the heat transfer in a partially blocked rod bundle. However, they did not consider the occurrence of fuel relocation and the corresponding effect on two-phase heat transfer. Some fragmented fuel particles located above the ballooned region may drop into the enlarged volume of the balloon. Accordingly, the fuel relocation brings in a local power increase in the ballooned region. The present study’s objective is to investigate the effect of the fuel relocation on the reflood under a LBLOCA condition. Toward this end, experiments were performed in a 5 × 5 partially-blocked rod bundle. Two power profiles were tested: one is a typical cosine shape and the other is the modified shape considering the effect of the fuel relocation. For a typical power shape, the peak temperature in the ballooned rods was lower than that in the intact rods. On the other hand, for the modified power shape, the peak temperature in the ballooned rods was higher than that in the intact rods. Numerical simulations were also performed using the MARS code. The tendencies of the peak clad temperatures were well predicted.

  15. Downflow film boiling in a rod bundle at low pressure

    International Nuclear Information System (INIS)

    Hochreiter, L.E.; Rosal, E.R.; Fayfich, R.R.

    1978-01-01

    A series of low pressure downflow film boiling heat transfer experiments were conducted in a 14-foot (4.27 m) long electrically heater rod bundle containing 336 heater rods. The resulting data was compared with the Dougall-Rohsenow dispersed flow film boiling correlation. The data was found to lie below this correlation as the quality was increased. It is believed that buoyancy effects decreased the heat transfer in downflow film boiling. (author)

  16. CFD investigation of vertical rod bundles of supercritical water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Shang Zhi

    2009-01-01

    The commercial CFD code STAR-CD v4.02 is used as the numerical simulation tool for the supercritical water-cooled nuclear reactor (SCWR). The numerical simulation is based on the real full 3D rod bundles' geometry of the nuclear reactors. For satisfying the near-wall resolution of y + ≤ 1, the structure mesh with the stretched fine mesh near wall is employed. The validation of the numerical simulation for mesh generation strategy and the turbulence model for the heat transfer of supercritical water is carried out to compare with 3D tube experiments. After the validation, the same mesh generation strategy and the turbulence model are employed to study three types of the geometry frame of the real rod bundles. Through the numerical investigations, it is found that the different arrangement of the rod bundles will induce the different temperature distribution at the rods' walls. The wall temperature distributions are non-uniform along the wall and the values depend on the geometry frame. At the same flow conditions, downward flow gets higher wall temperature than upward flow. The hexagon geometry frame has the smallest wall temperature difference comparing with the others. The heat transfer is controlled by P/D ratio of the bundles.

  17. Rod displacement measurements by x-ray CT and its impact on thermal-hydraulics in tight-lattice rod bundle (Joint research)

    International Nuclear Information System (INIS)

    Mitsutake, Toru; Misawa, Takeharu; Kureta, Masatoshi; Akimoto, Hajime

    2005-06-01

    In tight-lattice simulated rod bundles with about 1 mm gap between rods, a rod displacement might affect thermal-hydraulic characteristics since the displacement has a strong impact on the flow area change along the heated section. It should be important to estimate how large the rod position displacement could quantitatively affect critical power for the tight-lattice rod bundle from the point of improvement of prediction capability of subchannel analysis. In the present study, the inside-structure observation of the simulated seven-rod bundle of Reduced Moderation Water Reactor (RMWR) was made through the whole length of the test assembly. Based on the measured rod position data, the relation between the rod position displacement and the heat transfer characteristics was investigated experimentally and through the two kinds of subchannel analysis, the nominal rod position case and the measured rod position case, the effect on the predicted critical power was estimated. The high-energy X-ray computer tomograph (CT) of Fuels Monitoring Facilities (FMF) at the O-arai Engineering Center in Japan Nuclear Cycle Institute (JNC) was applied for the inside-structure observation of the test assembly. The CT view of the cross sections within the test assembly assured the hexagonal rod position arrangement was almost the same as expected by design. The measured data with the X-ray CT facility showed that all rod displacements were small, 0.5 millimeters at maximum and 0.2 millimeters in average. In the heat transfer experiments for the seven-rod bundle, the boiling transition (BT) position and the rod surface temperature behavior was measured. All thermocouples on the center rod downstream from the BT-onset axial height showed almost simultaneous temperature increase due to BT. And the thermocouples located on the same axial heights showed quite similar time-variation behaviors in the vapor cooling heat transfer regime. These results demonstrated the effect of the

  18. Local thermal-hydraulic behaviour in tight 7-rod bundles

    International Nuclear Information System (INIS)

    Cheng, X.; Yu, Y.Q.

    2009-01-01

    Advanced water-cooled reactor concepts with tight lattices have been proposed worldwide to improve the fuel utilization and the economic competitiveness. In the present work, experimental investigations were performed on thermal-hydraulic behaviour in tight hexagonal 7-rod bundles under both single-phase and two-phase conditions. Freon-12 was used as working fluid due to its convenient operating parameters. Tests were carried out under both single-phase and two-phase flow conditions. Rod surface temperatures are measured at a fixed axial elevation and in various circumferential positions. Test data with different radial power distributions are analyzed. Measured surface temperatures of unheated rods are used for the assessment of and comparison with numerical codes. In addition, numerical simulation using sub-channel analysis code MATRA and the computational fluid dynamics (CFD) code ANSYS-10 is carried out to understand the experimental data and to assess the validity of these codes in the prediction of flow and heat transfer behaviour in tight rod bundle geometries. Numerical results are compared with experimental data. A good agreement between the measured temperatures on the unheated rod surface and the CFD calculation is obtained. Both sub-channel analysis and CFD calculation indicates that the turbulent mixing in the tight rod bundle is significantly stronger than that computed with a well established correlation.

  19. The Preliminary Study for Numerical Computation of 37 Rod Bundle in CANDU Reactor

    International Nuclear Information System (INIS)

    Jeon, Yu Mi; Bae, Jun Ho; Park, Joo Hwan

    2010-01-01

    A typical CANDU 6 fuel bundle consists of 37 fuel rods supported by two endplates and separated by spacer pads at various locations. In addition, the bearing pads are brazed to each outer fuel rod with the aim of reducing the contact area between the fuel bundle and the pressure tube. Although the recent progress of CFD methods has provided opportunities for computing the thermal-hydraulic phenomena inside of a fuel channel, it is yet impossible to reflect the detailed shape of rod bundle on the numerical computation due to a lot of computing mesh and memory capacity. Hence, the previous studies conducted a numerical computation for smooth channels without considering spacers, bearing pads. But, it is well known that these components are an important factor to predict the pressure drop and heat transfer rate in a channel. In this study, the new computational method is proposed to solve the complex geometry such as a fuel rod bundle. In front of applying the method to the problem of 37 rod bundle, the validity and the accuracy of the method are tested by applying the method to the simple geometry. Based on the present result, the calculation for the fully shaped 37-rod bundle is scheduled for the future works

  20. Experimental study of laminar mixed convection in a rod bundle with mixing vane spacer grids

    Energy Technology Data Exchange (ETDEWEB)

    Mohanta, Lokanath, E-mail: lxm971@psu.edu [Department of Mechanical and Nuclear Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Cheung, Fan-Bill [Department of Mechanical and Nuclear Engineering, Pennsylvania State University, University Park, PA 16802 (United States); Bajorek, Stephen M.; Tien, Kirk; Hoxie, Chris L. [Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 (United States)

    2017-02-15

    Highlights: • Investigated the heat transfer during mixed laminar convection in a rod bundle with linearly varying heat flux. • The Nusselt number increases downstream of the inlet with increasing Richardson number. • Developed an enhancement factor to account for the effects of mixed convection over the forced laminar heat transfer. - Abstract: Heat transfer by mixed convection in a rod bundle occurs when convection is affected by both the buoyancy and inertial forces. Mixed convection can be assumed when the Richardson number (Ri = Gr/Re{sup 2}) is on the order of unity, indicating that both forced and natural convection are important contributors to heat transfer. In the present study, data obtained from the Rod Bundle Heat Transfer (RBHT) facility was used to determine the heat transfer coefficient in the mixed convection regime, which was found to be significantly larger than those expected assuming purely forced convection based on the inlet flow rate. The inlet Reynolds (Re) number for the tests ranged from 500 to 1300, while the Grashof (Gr) number varied from 1.5 × 10{sup 5} to 3.8 × 10{sup 6} yielding 0.25 < Ri < 4.3. Using results from RBHT test along with the correlation from the FLECHT-SEASET test program for laminar forced convection, a new correlation ​is proposed for mixed convection in a rod bundle. The new correlation accounts for the enhancement of heat transfer relative to laminar forced convection.

  1. Data report of a tight-lattice rod bundle thermal-hydraulic tests (1). Base case test using 37-rod bundle simulated water-cooled breeder reactor (Contract research)

    International Nuclear Information System (INIS)

    Kureta, Masatoshi; Tamai, Hidesada; Liu, Wei; Akimoto, Hajime; Sato, Takashi; Watanabe, Hironori; Ohnuki, Akira

    2006-03-01

    Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests to realize essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor featured by a high breeding ratio and super high burn-up by reducing the core water volume in water-cooled reactor. The tests are performing to make clear the fundamental subjects related to the boiling transition (BT) (Subjects: BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0mm) and Rod-bowing one)). In the present report, we summarize the test results from the base case test section. The thermal-hydraulic characteristics using the large scale test section were obtained for the critical power, the pressure drop and the wall heat transfer under a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Effects of local peaking factor on the critical power were also obtained. (author)

  2. CFD evaluation of turbulence model on heat transfer in 5 × 5 rod bundles

    International Nuclear Information System (INIS)

    Chao Yanmeng; Yang Lixin; Zhang Yuxiang; Pang Zhengzheng

    2014-01-01

    Different turbulence models may lead to different results when analyzing fuel assemblies using computational fluid dynamics (CFD) method. In this paper, a 5 × 5 rod bundle model was built to analyze the relationship between flow and heat transfer. The pressure drop and Nu were calculated using ANSYS CFX. Three factors evaluating swirling flow and cross-flow were used to analyze the inner relationship between flow field and heat transfer. The performances of various turbulence models, including eddy viscosity model and Reynold stress model, were evaluated. The comparison between numerical and similar experimental results indicates that Reynold stress model is more appropriate for modeling flow features and heat transfer in spacer grids discussed in this paper. (authors)

  3. Flow in rod bundles

    International Nuclear Information System (INIS)

    Hazi, G.; Mayer, G.

    2005-01-01

    For power upgrading VVER-440 reactors we need to know exactly how the temperature measured by the thermocouples is related to the average outlet temperature of the fuel assemblies. Accordingly, detailed knowledge on mixing process in the rod bundles and in the fuel assembly head have great importance. Here we study the hydrodynamics of rod bundles based on the results of direct numerical and large eddy simulation of flows in subchannels. It is shown that secondary flow and flow pulsation phenomena can be observed using both methodologies. Some consequences of these observations are briefly discussed. (author)

  4. A Secondary Flow Effect on the Heat and Mass Transfer Processes in the Finned Rod Bundles of Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Dunaitsev

    2017-01-01

    Full Text Available In nuclear power engineering a need to justify an operability of products and their components is of great importance. In high-temperature gas reactors, the critical element affecting the facility reliability is the fuel rod cladding, which in turn leads to the need to gain knowledge in the field of gas dynamics and heat transfer in the reactor core and to increase the detail of the calculation results. For the time being, calculations of reactor core are performed using the proven techniques of per-channel calculations, which show good representativeness and count rate. However, these techniques require additional experimental studies to describe correctly the inter-channel exchange, which, being taken into account, largely affects the pattern of the temperature fields in the region under consideration. Increasingly more relevant and demandable are numerical simulation methods of fluid and gas dynamics, as well as of heat exchange, which consist in the direct solution of the system of differential equations of mass balance, kinetic moment, and energy. Calculation of reactor cores or rod bundles according these techniques does not require additional experimental studies and allows us to obtain the local distributions of flow characteristics in the bundle and the flow characteristics that are hard to measure in the physical experiment.The article shows the calculation results and their analysis for an infinite rod lattice of the reactor core. The results were obtained by the technique of modelling one rod of a regular lattice using the periodic boundary conditions, followed by translating the results to the neighbouring rods. In channels of complex shape, there are secondary flows caused by changes in the channel geometry along the flow and directed across the main front of the flow. These secondary flows in the reactor cores with rods spaced by the winding wire lead to a redistribution of the coolant along the channel section, which in turn

  5. Data report of tight-lattice rod bundle thermal-hydraulic tests (2). Gap-width effect test using 37-rod bundle simulated water-cooled breeder reactor (Contract research)

    International Nuclear Information System (INIS)

    Tamai, Hidesada; Kureta, Masatoshi; Liu, Wei; Akimoto, Hajime; Sato, Takashi; Watanabe, Hironori; Ohnuki, Akira

    2006-11-01

    Japan Atomic Energy Agency has been performing tight-lattice rod bundle thermal-hydraulic tests to realize essential technologies for the technological and engineering feasibility of super high burn-up water-cooled breeder reactor featured by a high breeding ratio and super high burn-up by reducing the core water volume in water-cooled reactor. The tests are performing to make clear the fundamental subjects related to the boiling transition (BT) (Subjects: BT criteria under a highly tight-lattice rod bundle, effects of gap-width between rods and of rod-bowing) using 37-rod bundles (Base case test section (1.3mm gap-width), Two parameter effect test sections (Gap-width effect one (1.0mm) and Rod-bowing one)). In the present report, we summarize the test results from the gap-width effect test section. The thermal-hydraulic characteristics were obtained for the critical power under the steady-state and transient conditions, the pressure drop and the wall heat transfer within a wide range of pressure, flow rate, etc. including normal operational conditions of the designed reactor. Then the gap-width effects were also obtained from the comparison between the results using the base case test section and the gap-width effect one. (author)

  6. Critical heat flux under zero flow conditions in a vertical 3 X 3 rod bundle with a non-uniform axial heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Seok; Chun, Se Young; Moon, Sang Ki; Baek, Won Pil

    2003-11-01

    KAERI has performed an experimental study of water Critical Heat Flux (CHF) under zero flow conditions with a non-uniformly heated 3 by 3 rod bundle. Experimental conditions are in the range of a system pressure from 0.5 to 15.0 MPa and inlet water subcooling enthalpies from 67.5 to 351.5 kJ/kg. The test section used in the present experiments consisted of a vertical flow channel, upper and lower plenums, and a non-uniformly heated 3 by 3 rod bundle. The experimental results show that the CHFs in low-pressure conditions are somewhat scattered within a narrow range. As the system pressure increases, however, the CHFs show a consistent parametric trend. The CHFs occur in the upper region of the heated section, but the vertical distances of the detected CHFs from the bottom of the heated section are reduced as the system pressure increases. Even though the effects of the inlet water subcooling enthalpies and system pressure in the flooding CHF are relatively smaller than those of the flow boiling CHF, the CHF increases by increasing the inlet water subcooling enthalpies. Several existing correlations for the countercurrent flooding CHF based on Wallis's flooding correlation and Kutateladze's criterion for the onset of flooding are compared with the CHF data obtained in the present experiments to examine the applicability of the correlations.

  7. Experimental pressure drop and heat transfer in square array rod bundle for fusion-fission hybrid system

    Energy Technology Data Exchange (ETDEWEB)

    Shamim, J.A.; Bhowmik, P.K. [Seoul National Univ., Gwanak Gu, Seoul (Korea, Republic of); Suh, K.Y., E-mail: kysuh@snu.ac.kr [Seoul National Univ., Gwanak Gu, Seoul (Korea, Republic of); PhiloSophia Inc., Gwanak Gu, Seoul (Korea, Republic of)

    2014-07-01

    The effects of grid spacer flow restriction on pressure drop are evaluated experimentally for a wide range of flow rates. The results are compared against predictions by using most well known correlations. The convective heat transfer coefficients are evaluated using ANSYS 12.1 for a 3x3 rod bundle for pure water and alumina nanofluid. It is observed that the experimental pressure drop falls within 10%~20% of the predictions. Heat transfer of the 4% alumina nanofluid increases about 18% over pure water under the same inlet flow condition. (author)

  8. Experimental pressure drop and heat transfer in square array rod bundle for fusion-fission hybrid system

    International Nuclear Information System (INIS)

    Shamim, J.A.; Bhowmik, P.K.; Suh, K.Y.

    2014-01-01

    The effects of grid spacer flow restriction on pressure drop are evaluated experimentally for a wide range of flow rates. The results are compared against predictions by using most well known correlations. The convective heat transfer coefficients are evaluated using ANSYS 12.1 for a 3x3 rod bundle for pure water and alumina nanofluid. It is observed that the experimental pressure drop falls within 10%~20% of the predictions. Heat transfer of the 4% alumina nanofluid increases about 18% over pure water under the same inlet flow condition. (author)

  9. Numerical investigation of heat transfer in upward flows of supercritical water in circular tubes and tight fuel rod bundles

    International Nuclear Information System (INIS)

    Yang Jue; Oka, Yoshiaki; Ishiwatari, Yuki; Liu Jie; Yoo, Jaewoon

    2007-01-01

    Heat transfer in upward flows of supercritical water in circular tubes and in tight fuel rod bundles is numerically investigated by using the commercial CFD code STAR-CD 3.24. The objective is to have more understandings about the phenomena happening in supercritical water and for designs of supercritical water cooled reactors. Some turbulence models are selected to carry out numerical simulations and the results are compared with experimental data and other correlations to find suitable models to predict heat transfer in supercritical water. The comparisons are not only in the low bulk temperature region, but also in the high bulk temperature region. The two-layer model (Hassid and Poreh) gives a better prediction to the heat transfer than other models, and the standard k-ε high Re model with the standard wall function also shows an acceptable predicting capability. Three-dimensional simulations are carried out in sub-channels of tight square lattice and triangular lattice fuel rod bundles at supercritical pressure. Results show that there is a strong non-uniformity of the circumferential distribution of the cladding surface temperature, in the square lattice bundle with a small pitch-to-diameter ratio (P/D). However, it does not occur in the triangular lattice bundle with a small P/D. It is found that this phenomenon is caused by the large non-uniformity of the flow area in the cross-section of sub-channels. Some improved designs are numerically studied and proved to be effective to avoid the large circumferential temperature gradient at the cladding surface

  10. A model for dispersed flow heat transfer in rod bundles during reflood

    International Nuclear Information System (INIS)

    Wong, S.

    1980-01-01

    The present model calculates the heat transfer characteristics of the non-equilibrium dispersed droplet flow regime above the quench front during reflood by solving simultaneously the wall-to-vapor interactions, wall-to-droplet interactions and vapor-to-droplet interactions by an iterative numerical method. The unique feature in the present study is various heat transfer mechanisms are combined in an overall energy balance equation, and the convective heat transfer to vapor is obtained by calculating the vapor temperature distributions at the heated walls. The reactor rod bundle geometry, axial variations of vapor temperature and flow properties, radiative heat transfers, and enhancement of heat transfer due to turbulence are considered carefully, so that the present model could be used to predict PWR (Pressurized Water Reactor) reflood heat transfers, and hence the fuel cladding wall temperature transients. In order to achieve closure of the problem formulations, the droplet size and its motion are determined from the FLECHT (Full Length Emergency Cooling Heat Transfer Program) low flooding rate series consine axial power shape test data. The model is then verified by comparing the heat transfer predictions with FLECHT low flooding rate series skewed axial power shape test data. Comparisons of predictions with data show satisfactory agreements

  11. Study of thermal hydraulic behavior of supercritical water flowing through fuel rod bundles

    International Nuclear Information System (INIS)

    Thakre, Sachin; Lakshmanan, S.P.; Kulkarni, Vinayak; Pandey, Manmohan

    2009-01-01

    Investigations on thermal-hydraulic behavior in Supercritical Water Reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community because of its potential to obtain high thermal efficiency and compact design. Present work deals with CFD analysis to study the flow and heat transfer behavior of supercritical water in 4 metre long 7-pin fuel bundle using commercial CFD package ANSYS CFX for single phase steady state conditions. Considering the symmetric conditions, 1/12th part of the fuel rod bundle is taken as a domain of analysis. RNG K-epsilon model with scalable wall functions is used for modeling the turbulence behavior. Constant heat flux boundary condition is applied at the fuel rod surface. IAPWS equations of state are used to compute thermo-physical properties of supercritical water. Sharp variations in its thermo-physical properties (specific heat, density) are observed near the pseudo-critical temperature causing sharp change in heat transfer coefficient. The pseudo-critical point initially appears in the gaps among heated fuel rods, and then spreads radially outward reaching the adiabatic wall as the flow goes downstream. The enthalpy gain in the centre of the channel is much higher than that in the wall region. Non-uniformity in the circumferential distribution of surface temperature and heat transfer coefficient is observed which is in agreement with published literature. Heat transfer coefficient is high on the rod surface near the tight region and decreases as the distance between rod surfaces increases. (author)

  12. Analytical prediction of turbulent friction factor for a rod bundle

    International Nuclear Information System (INIS)

    Bae, Jun Ho; Park, Joo Hwan

    2011-01-01

    An analytical calculation has been performed to predict the turbulent friction factor in a rod bundle. For each subchannel constituting a rod bundle, the geometry parameters are analytically derived by integrating the law of the wall over each subchannel with the consideration of a local shear stress distribution. The correlation equations for a local shear stress distribution are supplied from a numerical simulation for each subchannel. The explicit effect of a subchannel shape on the geometry parameter and the friction factor is reported. The friction factor of a corner subchannel converges to a constant value, while the friction factor of a central subchannel steadily increases with a rod distance ratio. The analysis for a rod bundle shows that the friction factor of a rod bundle is largely affected by the characteristics of each subchannel constituting a rod bundle. The present analytic calculations well predict the experimental results from the literature with rod bundles in circular, hexagonal, and square channels.

  13. Establishment and assessment of CHF data base for square-lattice rod bundles

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun; Seo, K. W.; Kim, K. K.; Zee, S. Q.

    2002-02-01

    A CHF data base is constructed for square-lattice rod bundles, and assessed with various existing CHF prediction models. The CHF data base consists of 10725 data points obtained from 147 test bundles with uniform axial power distributions and 29 test bundles with non-uniform axial power distributions. The local thermal-hydraulic conditions in the subchannels are calculated by employing a subchannel analysis code MATRA. The influence of turbulent mixing parameter on CHF is evaluated quantitatively for selected test bundles with representative cross sectional configurations. The performance of various CHF prediction models including empirical correlations for round tubes or rod bundles, theoretical DNB models such as sublayer dryout model and bubble crowding model, and CHF lookup table for round tubes, are assessed for the localized rod bundle CHF data base. In view of the analysis result, it reveals that the 1995 AECL-IPPE CHF lookup table method is one of promising models in the aspect of the prediction accuracy and the applicable range. As the result of analysis employing the CHF lookup table for 9113 data points with uniform axial heat profile, the mean and the standard deviation of P/M are calculated as 1.003 and 0.115 by HBM, 1.022 and 0.319 by DSM respectively

  14. Feasibility evaluation of x-ray imaging for measurement of fuel rod bowing in CFTL test bundles

    International Nuclear Information System (INIS)

    Baker, S.P.

    1980-06-01

    The Core Flow Test Loop (CFTL) is a high temperature, high pressure, out-of-reactor helium-circulating system. It is designed for detailed study of the thermomechanical performance, at prototypic steady-state and transient operating conditions, of electrically heated rods that simulate segments of core assemblies in the Gas-Cooled Fast Breeder reactor demonstration plant. Results are presented of a feasibility evaluation of x-ray imaging for making measurements of the displacement (bowing) of fuel rods in CFTL test bundles containing electrically heated rods. A mock-up of a representative CFTL test section consisting of a test bundle and associated piping was fabricated to assist in this evaluation

  15. Computational fluid dynamics (CFD) round robin benchmark for a pressurized water reactor (PWR) rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Shin K., E-mail: paengki1@tamu.edu; Hassan, Yassin A.

    2016-05-15

    Highlights: • The capabilities of steady RANS models were directly assessed for full axial scale experiment. • The importance of mesh and conjugate heat transfer was reaffirmed. • The rod inner-surface temperature was directly compared. • The steady RANS calculations showed a limitation in the prediction of circumferential distribution of the rod surface temperature. - Abstract: This study examined the capabilities and limitations of steady Reynolds-Averaged Navier–Stokes (RANS) approach for pressurized water reactor (PWR) rod bundle problems, based on the round robin benchmark of computational fluid dynamics (CFD) codes against the NESTOR experiment for a 5 × 5 rod bundle with typical split-type mixing vane grids (MVGs). The round robin exercise against the high-fidelity, broad-range (covering multi-spans and entire lateral domain) NESTOR experimental data for both the flow field and the rod temperatures enabled us to obtain important insights into CFD prediction and validation for the split-type MVG PWR rod bundle problem. It was found that the steady RANS turbulence models with wall function could reasonably predict two key variables for a rod bundle problem – grid span pressure loss and the rod surface temperature – once mesh (type, resolution, and configuration) was suitable and conjugate heat transfer was properly considered. However, they over-predicted the magnitude of the circumferential variation of the rod surface temperature and could not capture its peak azimuthal locations for a central rod in the wake of the MVG. These discrepancies in the rod surface temperature were probably because the steady RANS approach could not capture unsteady, large-scale cross-flow fluctuations and qualitative cross-flow pattern change due to the laterally confined test section. Based on this benchmarking study, lessons and recommendations about experimental methods as well as CFD methods were also provided for the future research.

  16. High-resolution flow structure measurements in a rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Ylönen, A. T.

    2013-07-01

    Flow behaviour inside a rod bundle has been an active research topic since the early days of the nuclear power industry. Of particular interest in previous studies have been topics such as flow mixing, two-phase flow structure and mapping of two-phase flow transitions. The optimisation of fuel element design can only be achieved by truly understanding the nature of flow. The ultimate goal in this research is to enhance the heat transfer and increase the critical heat flux, which would improve the fuel economy. A better understanding of the flow would also improve nuclear safety as departure from nucleate boiling (DNB) can be predicted more accurately. The motivation for the current project (SUBFLOW) was to increase knowledge of the complex flow phenomena inside a rod bundle. A dedicated sub-channel flow test facility was designed and constructed at the Paul Scherrer Institut (PSI), Villigen, Switzerland. An adiabatic test loop has an up-scaled (1:2.6) vertical fuel rod bundle model with a 4 × 4 geometry. For the very first time, the wire-mesh sensor measurement technique was implemented in a rod bundle as two 64×64 conductivity wire-mesh sensors were installed in the upper part of the test section. The measurement technique enables one to study single- and two-phase flow behaviour with high spatial and temporal resolution. The research topics addressed in this thesis cover a wide range of flow conditions with and without a spacer grid in a rod bundle. The experimental campaign was started by studying natural mixing of a passive scalar to characterise the development of turbulent diffusion in an injection sub-channel and, later on, cross-mixing between adjacent sub-channels. The results were also used in comparison with the in-house CFD code PSI-Boil that is being developed at PSI. The code could estimate the mixing inside the sub-channel and the transition to cross-mixing with a good accuracy. As a natural transition, the SUBFLOW experiments were continued by

  17. High-resolution flow structure measurements in a rod bundle

    International Nuclear Information System (INIS)

    Ylönen, A. T.

    2013-01-01

    Flow behaviour inside a rod bundle has been an active research topic since the early days of the nuclear power industry. Of particular interest in previous studies have been topics such as flow mixing, two-phase flow structure and mapping of two-phase flow transitions. The optimisation of fuel element design can only be achieved by truly understanding the nature of flow. The ultimate goal in this research is to enhance the heat transfer and increase the critical heat flux, which would improve the fuel economy. A better understanding of the flow would also improve nuclear safety as departure from nucleate boiling (DNB) can be predicted more accurately. The motivation for the current project (SUBFLOW) was to increase knowledge of the complex flow phenomena inside a rod bundle. A dedicated sub-channel flow test facility was designed and constructed at the Paul Scherrer Institut (PSI), Villigen, Switzerland. An adiabatic test loop has an up-scaled (1:2.6) vertical fuel rod bundle model with a 4 × 4 geometry. For the very first time, the wire-mesh sensor measurement technique was implemented in a rod bundle as two 64×64 conductivity wire-mesh sensors were installed in the upper part of the test section. The measurement technique enables one to study single- and two-phase flow behaviour with high spatial and temporal resolution. The research topics addressed in this thesis cover a wide range of flow conditions with and without a spacer grid in a rod bundle. The experimental campaign was started by studying natural mixing of a passive scalar to characterise the development of turbulent diffusion in an injection sub-channel and, later on, cross-mixing between adjacent sub-channels. The results were also used in comparison with the in-house CFD code PSI-Boil that is being developed at PSI. The code could estimate the mixing inside the sub-channel and the transition to cross-mixing with a good accuracy. As a natural transition, the SUBFLOW experiments were continued by

  18. A thermal-hydraulic code for transient analysis in a channel with a rod bundle

    International Nuclear Information System (INIS)

    Khodjaev, I.D.

    1995-01-01

    The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident

  19. A thermal-hydraulic code for transient analysis in a channel with a rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Khodjaev, I.D. [Research & Engineering Centre of Nuclear Plants Safety, Electrogorsk (Russian Federation)

    1995-09-01

    The paper contains the model of transient vapor-liquid flow in a channel with a rod bundle of core of a nuclear power plant. The computer code has been developed to predict dryout and post-dryout heat transfer in rod bundles of nuclear reactor core under loss-of-coolant accidents. Economizer, bubble, dispersed-annular and dispersed regimes are taken into account. The computer code provides a three-field representation of two-phase flow in the dispersed-annular regime. Continuous vapor, continuous liquid film and entrained liquid drops are three fields. For the description of dispersed flow regime two-temperatures and single-velocity model is used. Relative droplet motion is taken into account for the droplet-to-vapor heat transfer. The conservation equations for each of regimes are solved using an effective numerical technique. This technique makes it possible to determine distribution of the parameters of flows along the perimeter of fuel elements. Comparison of the calculated results with the experimental data shows that the computer code adequately describes complex processes in a channel with a rod bundle during accident.

  20. AgInCd control rod failure in the QUENCH-13 bundle test

    International Nuclear Information System (INIS)

    Sepold, L.; Lind, T.; Csordas, A. Pinter; Stegmaier, U.; Steinbrueck, M.; Stuckert, J.

    2009-01-01

    The QUENCH off-pile experiments performed at the Karlsruhe Research Center are to investigate the high-temperature behavior of Light Water Reactor (LWR) core materials under transient conditions and in particular the hydrogen source term resulting from the water injection into an uncovered LWR core. The typical LWR-type QUENCH test bundle, which is electrically heated, consists of 21 fuel rod simulators with a total length of approximately 2.5 m. The Zircaloy-4 rod claddings and the grid spacers are identical to those used in Pressurized Water Reactors (PWR) whereas the fuel is represented by ZrO 2 pellets. In the QUENCH-13 experiment the single unheated fuel rod simulator in the center of the test bundle was replaced by a PWR-type control rod. The QUENCH-13 experiment consisting of pre-oxidation, transient, and quench water injection at the bottom of the test section investigated the effect of an AgInCd/stainless steel/Zircaloy-4 control rod assembly on early-phase bundle degradation and on reflood behavior. Furthermore, in the frame of the EU 6th Framework Network of Excellence SARNET, release and transport of aerosols of a failed absorber rod were to be studied in QUENCH-13, which was accomplished with help of aerosol measurements performed by PSI-Switzerland and AEKI-Hungary. Control rod failure was initiated by eutectic interaction of steel cladding and Zircaloy-4 guide tube and was indicated at about 1415 K by axial peak absorber and bundle temperature responses and additionally by the on-line aerosol monitoring system. Significant releases of aerosols and melt relocation from the control rod were observed at an axial peak bundle temperature of 1650 K. At a maximum bundle temperature of 1820 K reflood from the bottom was initiated with cold water at a flooding rate of 52 g/s. There was no noticeable temperature escalation during quenching. This corresponds to the small amount of about 1 g in hydrogen production during the quench phase (compared to 42 g of H 2

  1. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    International Nuclear Information System (INIS)

    Korteniemi, V.; Haapalehto, T.; Puustinen, M.

    1995-01-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied

  2. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    Energy Technology Data Exchange (ETDEWEB)

    Korteniemi, V.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Puustinen, M. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied.

  3. Evaluation of CHF experimental data for non-square lattice 7-rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Yoo, Y. J.; Kim, K. K.; Zee, S. Q

    2001-01-01

    A series of CHF experiments are conducted for 7-rod hexagonal test bundles in order to investigate the CHF characteristics of self-sustained square finned (SSF) rod bundles. The experiments are performed in the freon-loop and water-loop located at IPPE in Russia, and 609 data of freon-12 and 229 data of water are obtained from 7 kinds of test bundles classified by the combination of heated length and axial/radial power distributions. As the result of the evaluation of four representative CHF correlations, the EPRI-1 correlation reveals good prediction capability for SSF test bundles. The inlet parameter CHF correlation suggested by IPPE calculates the mean and the standard deviation of P/M for uniformly heated test bundles as 1.002 and 0.049, respectively. In spite of its excellent accuracy, the correlation has a discontinuity point at the boundary between the low velocity and high velocity conditions. KAERI's inlet parameter correlation eliminates this defect by introducing the complete evaporation model at low velocity condition, and calculates the mean and standard deviation of P/M as 0.095 and 0.062 for uniformly heated 496 data points, respectively. The mean/standard deviation of local parameter CHF correlations suggested by IPPE and KAERI are evaluated as 1.023/0.178 and 1.002/0.158, respectively. The inlet parameter correlation developed from uniformly heated test bundles tends to under-predict CHF about 3% for axially non-uniformly heated test bundles. On the other hand, the local parameter correlation reveals large scattering of P/M, and requires re-optimization of the correlation for non-uniform axial power distributions. As the result of the analysis of experimental data, it reveals that the correction model of axial power shapes suggested by IPPE is applicable to the inlet parameter correlations. For the test bundle of radial non-uniform power distribution, the physically unexpected results are obtained at some experimental conditions. In addition

  4. To the problem of interchannel mixing of a coolant in finned-rod bundles

    International Nuclear Information System (INIS)

    Dzyubenko, B.V.

    1979-01-01

    Suggested is the methods of experimental data processing on interchannel mixing in finned-rod bundles, based on the use of P-theorem of similarity and dimension theory and physical considerations. The obtained general conclusions of dependence for calculation of effective diffusion coefficients and coefficients of interchannel mixing permit to lock the systems of differential equations, describina the flow of homogenized media and the flow in a real rod bundle correspondingly. The carried out analysis of the experimental data obtained by different authors, who applied the method of central rod heating during the investigation of interchannel change, permitted to find essential influence of the suggested similarity criteria on mixing coefficient, characterizing the effect of centrifugal forces on the flow and a reformed longitudinal coordinate

  5. Evaluation of droplet deposition in rod bundle

    International Nuclear Information System (INIS)

    Ji, W.; Gu, C.Y.; Anglart, H.

    1997-01-01

    Deposition model for droplets in gas droplet two-phase flow in rod bundle is developed in this work using the Lagrangian method. The model is evaluated in a 9-rod bundle geometry. The deposition coefficient in the bundle geometry are compared with that in round tube. The influences of the droplet size and gas mass flow rate on deposition coefficient are investigated. Furthermore, the droplet motion is studied in more detail by dividing the bundle channel into sub-channels. The results show that the overall deposition coefficient in the bundle geometry is close to that in the round tube with the diameter equal to the bundle hydraulic diameter. The calculated deposition coefficient is found to be higher for higher gas mass flux and smaller droplets. The study in the sub-channels show that the ratio between the local deposition coefficient for a sub-channel and the averaged value for the whole bundle is close to a constant value, deviations from the mean value for all the calculated cases being within the range of ±13%. (author)

  6. Absorber rod bundle actuator in a pressurized water nuclear reactor

    International Nuclear Information System (INIS)

    Martin, J.; Peletan, R.

    1984-01-01

    The invention concerns an absorber rod bundle actuator in a pressurized water reactor with spectral shift control. The device comprises two coaxial control bars. The inner bar is integral with the absorber rod bundle; it has an enlarged zone which acts as a proton under pressure difference across an annular seal which can be radially expanded, the pressure difference allowing to the absorber rod bundles actuating on the piston. When a pressure difference is applied, the seal expands radially by a sufficient amount to make sealing contact with the zone of larger diameter in the outer bar. The invention applies more particularly to reactors with spectral shift control using bundles of fertile rods [fr

  7. Thermohydraulic tests of 3x3-rod bundle maquette

    International Nuclear Information System (INIS)

    Ladeira, L.C.D.

    1986-10-01

    The results of a 3x3-rod bundle thermohydraulic research program, performed in the Thermohydraulics Laboratory of NUCLEBRAS' Nuclear Technology Development Center, are briefly described. This program included measurements of pressure drops in one and two-phase flows, heat transfer coefficients, mixing between interconnected subchannels in one-phase flow conditions and critical heat fluxes. The measurements covered the following parameter ranges: heat fluxes from zero to the critical values, pressure ranging from 1 to 15 ata, inlet temperature from 25 to 150 sup(0)C and flow rate from 20 to 300l/min. (author)

  8. The turbulent flow in rod bundles

    International Nuclear Information System (INIS)

    Moeller, S.V.

    1989-01-01

    Experimental studies have shown that the axial and azimuthal turbulence intensities in the gap regions of rod bundles increase strongly with decreasing rod spacing; the fluctuating velocities in the axial and azimuthal directions have a quasi-periodic behaviour. To determine the origin of this phenomenon, an its characteristics as a function of the geometry and the Reynolds number, an experimental investigation was performed on the turbulent in several rod bundles with different aspect ratios (P/D, W/D). Hot-wires and microsphones were used for the measurements of velocity and wall pressure fluctuations. The data were evaluated to obtain spectra as well as auto and cross correlations. Based on the results, a phenomenological model is presented to explain this phenomenon. By means of the model, the mass exchange between neighbouring subchannels is explained [pt

  9. Heat transfer experiments and correlations for natural and forced circulations of water in rod bundles at low Reynolds numbers

    International Nuclear Information System (INIS)

    Kim, Sung-Ho; El-Genk, Mohamed S.; Rubio, Reuben A.; Bryson, James W.; Foushee, Fabian C.

    1988-01-01

    Experimental heat transfer studies were conducted for fully developed forced and natural flows of water through seven uniformly heated rod bundles, triangularly arrayed with P/D = 1.25, 1.38, and 1.5. In forced circulation experiments, Re ranged from 80 to 50,000 and Pr from 3 to 8.5, while in natural circulation, Re varied from 260 to 2,000, and Ra q from 8 x 10 8 to 2.5 x 10 8 . The forced flow data fell into the two basic flow regimes: turbulent and laminar flow. At the transition between these regimes, Re, which varied from 2,200 for P/D = 1.25 to 5,500 for P/D = 1.5, increased linearly with P/D. The heat transfer data for turbulent flow was within ±15 percent of Weisman's correlation, which was developed for fully developed turbulent flow in rod bundles at Re > 25,000. The laminar flow data showed the dependence of Nu on Re to be weaker than that for turbulent flow, but the exponent of Re increased with P/D: Nu = A Re B Pr 1/3 , where A is equal to 1.061, 0.511, and 0.346 for P/D = 1.25, 1.38 and 1.5, respectively, and B is a linear function of P/D (B = 0.797 P/D - 0.656). Natural circulation data indicated that rod spacing only slightly affected heat transfer, and Nu increased proportionally to Ra 0.25 ; Nu = 0.272 Ra q 0.25 . The application of the results to SNL's ACRR indicated that although the core is cooled by natural convection, either the natural circulation correlation or the forced turbulent flow correlation can be used to accurately predict the single phase heat transfer coefficient in the ACRR. These results were concluded because of the high Rayleigh and Reynolds numbers in the ACRR. The ACRR operates near the boundary between mixed and forced turbulent flow regimes: consequently, achieving the high heat transfer coefficient was possible with natural circulation. (author)

  10. CFD modeling of secondary flows in fuel rod bundles

    International Nuclear Information System (INIS)

    Baglietto, Emilio; Ninokata, Hisashi

    2004-01-01

    An optimized non-linear eddy viscosity model is introduced, for calculations of detailed coolant velocity distribution in a tight lattice fuel bundle. The low Reynolds formulation has been optimized based on DNS data for channel flow. The non-linear stress-strain relationship has been modified in the coefficients to model the flow anisotropy, which causes the formation of turbulence driven secondary flows inside the bundle subchannels. Predictions of the model are first compared to experimental measurements of secondary flows in a triangularly arrayed rod bundle with p/d=1.3. Subsequently wall shear stress and velocity predictions are compared with different experimental data for a rod bundle with p/d=1.17. The model shows to be able to correctly reproduce the scale of the secondary motion, and to accurately reproduce both wall shear stress and velocity distributions inside the rod bundle subchannels. (author)

  11. Experimental observation of the droplet size change across a wet grid spacer in a 6 × 6 rod bundle

    International Nuclear Information System (INIS)

    Cho, Hyoung Kyu; Choi, Ki Yong; Cho, Seok; Song, Chul-Hwa

    2011-01-01

    Highlights: ► In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. ► The experiment was focused on the change of droplet size induced by a spacer grid in a rod bundle. ► The major measuring parameters of the experiment were the droplet size and velocity. ► This test provided the data on the change of the droplet size after collision with a wet grid spacer. - Abstract: During the reflood phase of a postulated loss of coolant accident in a nuclear reactor, entrainment of liquid droplets can occur at a quench front of reflooding water. It is widely recognized that the behavior of the entrained droplets crucially affects the reflood heat transfer phenomena by decreasing the superheated steam temperature and interacting with a rod bundle and spacer grids. For this reason, various experimental and numerical studies have been performed to examine droplet behavior such as the droplet size, velocity and droplet fraction inside a rod array. In this study, an experiment on the droplet behavior inside a heated rod bundle has been performed. The experiment was focused on the change of droplet size induced by a spacer grid in a rod bundle geometry, which results in the change of the interfacial heat transfer between droplets and superheated steam. A 6 × 6 rod bundle test facility in Korea Atomic Energy Research Institute was used for the experiment. Steam was supplied by an external boiler into the bottom of the test channel, and a droplet injection nozzle was equipped instead of simulating a quench front of reflooding water. The major measuring parameters of the experiment were the droplet size and velocity, which were measured by a high-speed camera and a digital image processing technique. A series of experiments were conducted with various flow conditions of a steam injection velocity, heater temperature, droplet size, and droplet flow rate. The experiments provided the data on the change of the Sauter mean diameter of

  12. Critical power experiment with a tight-lattice 37-rod bundle

    International Nuclear Information System (INIS)

    Kureta, Masatoshi; Tamai, Hidesada; Ohnuki, Akira; Sato, Takashi; Liu, Wei; Akimoto, Hajime

    2006-01-01

    Since most of critical power or CHF data have been collected in tube, annulus, or BWR geometries under BWR flow conditions, critical power data for highly tight and triangular lattice bundles under low mass velocity are indispensable for thermal-hydraulic design of Reduced-Moderation Water Reactor. Large-scale thermal-hydraulic experiments which use a basic 37-rod bundle test section (rod diameter: 13.0 mm, gap width between rods: 1.3 mm) were therefore carried out in this study within range of 2-9 MPa in pressure and 150-1,000 kg/(m 2 ·s) in mass velocity. Fundamental characteristics of boiling transition were investigated through effects of flow parameter on critical power and those of rod number. It was confirmed that the fundamental characteristics in 37-rod bundle are similar to those in 7-rod bundle and in case of the BWR geometry. The results of the transverse non-uniform power distribution test and subchannel analysis suggest that the critical power becomes higher when the transverse local quality distribution closes to uniform. (author)

  13. Relative desorption of boiling crisis in rod bundles

    International Nuclear Information System (INIS)

    Bobkov, V.P.

    1997-01-01

    Results of describing critical heat fluxes rod bundles are presented on base of applying a generalization of the available massive of data on CHF in spherical tubes, performed on the base of a new model, developed by the physics and Power Institute specialists, as well as on the base of results of analysing comprehensive experimental material accumulated in the data bank of the Thermophysical Data Center of the PPI Ratios, allowing one to predict the values of the critical heat flux in a wide range of mode and geometry parameters under energy release with cross section variations and cross section geometry distortion are presented

  14. Rod-bundle transient-film boiling of high-pressure water in the liquid-deficient regime

    International Nuclear Information System (INIS)

    Morris, D.G.; Mullins, C.B.; Yoder, G.L.

    1982-01-01

    Results are reported from a recent experiment investigating dispersed flow film boiling of high pressure water in upflow through a rod bundle. The data, obtained under mildly transient conditions, are used to assess correlations currently used to predict heat transfer in these circumstances. In light of the scarcity of similar data, the data should prove useful in the development and assessment of new heat transfer models. The experiment was conducted at the Oak Ridge National Laboratory in the Thermal-Hydraulic Test Facility, a highly instrumented, non-nuclear, pressurized-water loop containing 64, 3.66-m (12-ft) long rods (of which 60 are electrically heated). The rods are arranged in a square array typical of 17 x 17 fuel rod assemblies in late generation PWRs. Data were collected over typical reactor blowdown parameter ranges

  15. 16-rod-bundle: Irradiation in the MZFR and post-irradiation examinations

    International Nuclear Information System (INIS)

    Manzel, R.

    1979-04-01

    In the course of the irradiation of a 16-rod prototype bundle, the basis has been established for the irradiation of experimental fuel assemblies containing full-length PWR fuel rods in standard positions of the MZFR. The prototype bundle was discharged after an irradiation time of 284 full power days and a burnup of 11400 MWd/tU. The overall performance of the prototype bundle was highly satisfactory. Detailed post-irradiation examinations confirmed the good conditions of bundle structures and fuel rods. (orig.) [de

  16. SCADOP: Phenomenological modeling of dryout in nuclear fuel rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Dasgupta, Arnab, E-mail: arnie@barc.gov.in; Chandraker, D.K., E-mail: dineshkc@barc.gov.in; Vijayan, P.K., E-mail: vijayanp@barc.gov.in

    2015-11-15

    Highlights: • Phenomenological model for annular flow dryout is presented. • The model evaluates initial entrained fraction using a new methodology. • The history effect in annular flow is predicted and validated. • Rod bundle dryout is predicted using subchannel methodology. • Model is validated against experimental dryout data in tubes and rod bundles. - Abstract: Analysis and prediction of dryout is of important consequence to safety of nuclear fuel clusters of boiling water type of reactors. Traditionally, experimental correlations are used for dryout predictions. Since these correlations are based on operating parameters and do not aim to model the underlying phenomena, there has been a proliferation of the correlations, each catering to some specific bundle geometry under a specific set of operating conditions. Moreover, such experiments are extremely costly. In general, changes in tested bundle geometry for improvement in thermal-hydraulic performance would require re-experimentation. Understanding and modeling the basic processes leading to dryout in flow boiling thus has great incentive. Such a model has the ability to predict dryout in any rod bundle geometry, unlike the operating parameter based correlation approach. Thus more informed experiments can be carried out. A good model can, reduce the number of experiments required during the iterations in bundle design. In this paper, a phenomenological model as indicated above is presented. The model incorporates a new methodology to estimate the Initial Entrained Fraction (IEF), i.e., entrained fraction at the onset of annular flow. The incorporation of this new methodology is important since IEF is often assumed ad-hoc and sometimes also used as a parameter to tune the model predictions to experimental data. It is highlighted that IEF may be low under certain conditions against the general perception of a high IEF due to influence of churn flow. It is shown that the same phenomenological model is

  17. PHEBUS/test-218, Behaviour of a Fuel Rod Bundle during a Large Break LOCA Transient with a two Peaks Temperature History

    International Nuclear Information System (INIS)

    1987-01-01

    1 - Description of test facility: PHEBUS test facility operated at CEA Research Center Cadarache consists of a pressurized circuit involving pumps, heat exchangers and a blowdown tank - 25 nuclear fuel rod bundle, coupled to a separate driver core; - active length 0.8 m, cosine axial power profile; - pressurized and un-pressurized fuel rods; - controlled cooling conditions at the bundle inlet (blowdown, refill and reflood period); - de-pressurized test rig volume 0.22 m 3 . The following 'as measured' boundary conditions (B.C.) were offered to participants as options with decreasing challenge to their analytical approach: Boundary conditions B.C.0: - full thermal-hydraulic analysis of PHEBUS test rig (was not recommended). Boundary conditions B.C.1: - thermal power level of fuel bundle; - fluid inlet conditions to bundle section. Boundary conditions B.C.2: - local cladding temperatures of rods; - heat transfer coefficients. Boundary conditions B.C.3: - cladding temperatures of rods; - internal pressure of rods. 2 - Description of test: Post-test investigation into the response of a nuclear fuel bundle to a large break loss of coolant accident with respect to - local fuel temperatures, - cladding strain at the time of burst, - time to burst and under given thermal-hydraulic boundary conditions of PHEBUS-test 218

  18. Full scale mock-up tests for rod bundle thermal-hydraulics in Japan

    International Nuclear Information System (INIS)

    Sugawara, S.

    1995-01-01

    This poster describes tests aimed at development and validation of principal design methodology of rod bundle thermal-hydraulics correlations. The works are based on domestic data base using the full-scale mock-up test facilities. The scope of the tests comprises DNB heat flux, transient DNB heat flux, post DNB heat transfer, pressure drop and void distribution. The works have been performed under collaboration among electric facilities, NPP vendors, universities, governmental corporations. 1 tab., 14 figs

  19. Comparisons of numerical simulations with ASTRID code against experimental results in rod bundle geometry for boiling flows

    International Nuclear Information System (INIS)

    Larrauri, D.; Briere, E.

    1997-12-01

    After different validation simulations of flows through cylindrical and annular channels, a subcooled boiling flow through a rod bundle has been simulated with ASTRID Steam-Water of software. The experiment simulated is called Poseidon. It is a vertical rectangular channel with three heating rods inside. The thermohydraulic conditions of the simulated flow were close to the DNB conditions. The simulation results were analysed and compared against the available measurements of liquid and wall temperatures. ASTRID Steam-Water produced satisfactory results. The wall and the liquid temperatures were well predicted in the different parts of the flow. The void fraction reached 40 % in the vicinity of the heating rods. The distribution of the different calculated variables showed that a three-dimensional simulation gives essential information for the analysis of the physical phenomena involved in this kind of flow. The good results obtained in Poseidon geometry will encourage future rod bundle flow simulations and analyses with ASTRID Steam-Water code. (author)

  20. Numerical investigation on practicability of reducing MCST by using grid spacer in a tight rod bundle

    International Nuclear Information System (INIS)

    Zhu, Xiaojing; Morooka, Shinichi; Oka, Yoshiaki

    2014-01-01

    Highlights: • Standard grid spacer design causes decreased heat transfer in a tight rod bundle. • Heat transfer is greatly enhanced by flow-enhancing features. • Swirling flow adversely affects the heat transfer downstream of grid spacer. • Enhanced heat transfer by existing grid spacer is limited in a short region. • Improved grid spacer can effectively reduce MCST. - Abstract: The numerical investigation was carried out to reveal the practicability of reducing the maximum cladding surface temperature (MCST) within the inner sub-channel of a tight, hexagon rod bundle using commercial CFD code STAR CCM+ 6.04. The special heat transfer and pressure drop characteristics caused by four existing grid spacer designs were discussed in detail by analyzing the effects of grid strap length, different flow enhancing features and different Reynolds numbers. It was found that the local heat transfer within the grid strap is greatly enhanced due to the raised flow velocity. Both the standard grid spacer and the grid spacer with split-vanes cause decreased heat transfer in the downstream region. The friction drag is very influential in the tight rod bundle and can eliminate the positive effect of flow blockage on the heat transfer performance. The grid spacer with flow blockage discs induces relatively good heat transfer performance and higher pressure drop within sub-channels, indicating a tradeoff between the heat transfer augmentation and the pressure drop. The combination of multiple existing grid spacers can reduce the MCST to a certain level, but the corresponding disadvantages cannot be ignored. The improved grid spacer design was proposed based on the overall considerations of heat transfer and pressure drop characteristics and has been proved more suitable to widely reduce MCST for SCWR than any other grid spacer designs involved in present study

  1. CFD analysis of flow and heat transfer in Canadian supercritical water reactor bundle

    International Nuclear Information System (INIS)

    Podila, K.; Rao, Y.F.

    2015-01-01

    Highlights: • Flow and heat transfer in SCWR fuel bundle design by AECL is studied using CFD. • Bare-rod bundle geometry is tested at 23.5, 25 and 28 MPa using STAR-CCM+ code. • SST k–ω low-Re model was used to study occurrence of heat transfer deterioration. - Abstract: Within the Gen-IV International Forum, AECL is leading the effort in developing a conceptual design for the Canadian SCWR. AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for occurrence of HTD in the supercritical bundle flows. In the current investigation, bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The low-Reynolds number modification of SST k–ω turbulence model along with y + < 1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5 MPa)

  2. PWR FLECHT SEASET 21-rod bundle flow blockage task. Task plan report. FLECHT SEASET Program report No. 5

    International Nuclear Information System (INIS)

    Hochreiter, L.E.; Basel, R.A.; Dennis, R.J.; Lee, N.; Massie, H.W. Jr.; Loftus, M.J.; Rosal, E.R.; Valkovic, M.M.

    1980-10-01

    This report presents a descriptive plan of tests for the 21-Rod Bundle Flow Blockage Task of the Full-Length Emergency Cooling Heat Transfer Separate Effects and Systems Effects Test Program (FLECHT SEASET). This task will consist of forced and gravity reflooding tests utilizing electrical heater rods to simulate PWR nuclear core fuel rod arrays. All tests will be performed with a cosine axial power profile. These tests are planned to be used to determine effects of various flow blockage configurations (shapes and distributions) on reflooding behavior, to aid in development/assessment of computational models in predicting reflooding behavior of flow blockage configurations, and to screen flow blockage configurations for future 161-rod flow blockage bundle tests

  3. Temperature escalation in PWR fuel rod simulator bundles due to the Zircaloy/steam reaction: Test ESBU-2A

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Wallenfels, K.P.; Peck, S.O.

    1984-07-01

    This report describes the test conduct and results of the bundle test ESBU-2A, which was run to investigate the temperature escalation of zircaloy clad fuel rods. This investigation of temperature escalation is part of a series of out-of-pile experiments, performed within the framework of the PNS Severe Fuel Damage Program. The test bundle was of a 3 x 3 array of fuel rod simulators with a 0.4 m heated length. The fuel rod simulators were electrically heated and consisted of tungsten heaters, UO 2 annular pellets, and zircaloy cladding. A nominal steam flow of 0.7 g/s was inlet to the bundle. The bundle was surrounded by a zircaloy shroud which was insulated with ZrO 2 fiber ceramic wrap. The initial heatup rate of the bundle was 0.4 0 C/s. The temperature escalation began at the 255 mm elevation after 1200 0 C had been reached. At this elevation, the measured peak temperature was limited to 1500 0 C. It was concluded from different thermocouple results, that induced by this first escalation melt was formed in the lower part of the bundle. Consequently, the escalation in the lower part must be much higher, at least up to the melting temperature of zircaloy. Due to the failure in the steam production system, steam starvation in the upper region may explain the beginning of the escalation at the 255 mm elevation. The maximum temperature reached was 2175 0 C on the center rod at the end of the test. The unregularities in the steam supply may be the reason for less oxidation than expected. (orig./GL) [de

  4. Forced, combined and natural convections of water in a vertical nine-rod bundle with a square lattice and P/C = 1.5

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Su, Bingjing; Guo, Zhanxiong

    1992-01-01

    Heat transfer correlations are developed for forced turbulent and laminar, combined, and natural convections of water in a uniformly heated, square arranged, nine-rod bundle having a P/D ratio of 1.5. In all correlations, the heated equivalent diameter is used in all the dimensionless quantities, and the water physical properties are evaluated at the water bulk temperature. In the experiments, Re is varied from 300 to 2.5 X 10 4 , Pr from 4 to 9, Ra q from 3 x 10 6 to 3 x 10 8 for natural convection and from 5 x 10 7 to 7 , 10 8 for combined convection, and Ri from 0.04 to 100. In both upflow and downflow experiments, the transition from forced turbulent to forced laminar convection occurs at Re T = 6,700; while the transition from forced laminar to buoyancy assisted combined convection occurs at Ri = 2.0. Results show that the rod arrangement in the bundle has little effect on the values of Nu in the forced and natural convection regimes. In general, Nu values for the square arranged rod bundle are less than 8% higher and less than 10% lower than those for a triangularly arranged rod bundle in the forced and natural convection regimes, respectively. 16 refs., 7 figs

  5. Experimental investigation of cooling by top spray and bottom flooding of a simulated 64 rod bundle for a BWR. Pt. 2. Main experiment with modified test section

    International Nuclear Information System (INIS)

    Nilsson, L.; Gustafson, L.; Harju, R.

    1978-06-01

    The cooling of an electrically heated, full scale 64-rod bundle has been investigated under simulated emergency core cooling conditions. Emphasis was laid on measurements of rod cladding and canister temperatures. By means of difference pressure measurements the levels in bundle, by-pass and downcomer could be estimated and thus the effective reflooding velocity. The test section was modified compared to the pre-tests, in order to improve system effects simulation. A new rod bundle was installed including a hollow, water, rod and 63 indirectly heated rods. Parameter effects of coolant mass flow rate and distribution, initial cladding temperature, pressure and power were studied. The effect of the way the test section was vented was also investigated and turned out to be very significant. (author)

  6. Critical heat flux in tubes and tight hexagonal rod lattices

    International Nuclear Information System (INIS)

    Erbacher, F.J.; Cheng Xu; Zeggel, W.

    1994-01-01

    The critical heat flux (CHF) in small-diameter tubes and in tight hexagonal 7-rod and 37-rod bundles was investigated in the KRISTA test facility, using Freon 12 as the working fluid. The measurements in tubes showed that the influence of the tube diameter on CHF cannot be described as suggested by earlier publications with sufficient accuracy. CHF in bundles is lower than in tubes under comparable conditions. The influence of spacers (grid spacers, wire wraps) on CHF was found to be governed by local steam qualities. A comparison of the test results with some CHF prediction methods showed that the look-up table method reproduces the test results in circular tubes most accurately. Combined with CHF look-up tables, subchannel analysis and Ahmad's fluid-to-fluid scaling law, Freon experiments have proven to be a suitable tool for CHF prediction in water-cooled rod bundles. (orig.) [de

  7. A study on the thermal hydraulics in rod bundles

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Yang, Sun Kyu

    1989-03-01

    In order to improve the thermal hydraulic characteristics of the nuclear reactor core, it is necessary to obtain better understanding of the coolant flow and the enthalpy distribution in complex rod bundle geometries. The purpose of this report is to obtain a comprehensive survey on the thermal hydraulic in rod bundles from both experimental and numerical point of view. From references on experimental study, measurement methods and results of the flow velocity and the pressure drop in the subchannels of rod bundles are expressed. The microscopic flow characteristics of the subchannels and spacer grid effect on the flow structure are described. Physical phenomena and measurement methods of the secondary flow are also described. From references on the numerical study, general numerical methods are expressed. Numerical studies on the laminar flow and turbulent flow such as 1-equation and 2-equation model are reviewed.(Author)

  8. Pressure loss in two-phase flow through a microchannel rod bundle

    International Nuclear Information System (INIS)

    Smith, A.C.; Hamm, L.L.; Qureshi, Z.; Steeper, T.J.

    1998-01-01

    The purpose of the microchannel rod bundle two-phase flow test described here was to provide data for benchmarking safety analyses for the accelerator production of tritium (APT). The objective was to obtain pressure loss data for a typical accelerator target rod bundle over a wide range of two-phase flow conditions. The test rod bundle assembly was fabricated for single-phase pressure drop tests conducted at Los Alamos National Laboratory (LANL) and subsequently used for the two-phase flow testing described here. The results for a typical case are given. These results fall generally in the slug flow regime for the horizontal flow results of Fukano and Kariyasaki for a 1.0-mm circular channel. Fukano and Kariyasaki found that surface tension effects were dominant in the 1-mm channel and report no churn regime. The results were also compared with the flow regime maps given by Triplett et al. for flow in discrete microchannels. Triplett employed both circular and trapezoidal channels, the latter to approximate the rod bundle interstitial flow channel shape. It was found that the rod bundle flow fell across the slug-to-churn flow regime transition reported by Triplett. This is consistent with the expectation that cross flow among channels would result in turbulent mixing and would suppress the formation of large discrete bubbles

  9. Calculation of Heat-Bearing Agent’s Steady Flow in Fuel Bundle

    Science.gov (United States)

    Amosova, E. V.; Guba, G. G.

    2017-11-01

    This paper introduces the result of studying the heat exchange in the fuel bundle of the nuclear reactor’s fuel magazine. The article considers the fuel bundle of the infinite number of fuel elements, fuel elements are considered in the checkerboard fashion (at the tops of a regular triangle a fuel element is a plain round rod. The inhomogeneity of volume energy release in the rod forms the inhomogeneity of temperature and velocity fields, and pressure. Computational methods for studying hydrodynamics in magazines and cores with rod-shape fuel elements are based on a significant simplification of the problem: using basic (averaged) equations, isobaric section hypothesis, porous body model, etc. This could be explained by the complexity of math description of the three-dimensional fluid flow in the multi-connected area with the transfer coefficient anisotropy, curved boundaries and technical computation difficulties. Thus, calculative studying suggests itself as promising and important. There was developed a method for calculating the heat-mass exchange processes of inter-channel fuel element motions, which allows considering the contribution of natural convection to the heat-mass exchange based on the Navier-Stokes equations and Boussinesq approximation.

  10. Development of design technology on thermal-hydraulic performance in tight-lattice rod bundles. II-rod bowing effect on boiling transition

    International Nuclear Information System (INIS)

    Liu, Wei; Tamai, Hidesada; Kureta, Masatoshi; Ohnuki, Akira; Takase, Kazuyuki; Akimoto, Hajime

    2007-01-01

    A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R and D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we will describe the critical power characteristics in a 37-rod tight-lattice bundle with rod-bowing under both steady and transient states. It is observed that no matter it is run under a steady or a transient state, boiling transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle. Steady critical power increases monotonically with the increase of mass velocity, with the decrease of inlet water temperature and with the decrease of exit pressure. These trends are same as those in the base case test without rod-bowing. The steady critical power with rod-bowing is about 10% lower than that without rod-bowing. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transitions are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with TRAC-BF1 code. The TRAC-BF1 code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time within the accuracy of critical power correlation. Traditional quasi - steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight lattice bundle with rod - bowing. (author)

  11. Theoretical and experimental investigations of CHF in round tubes and rod bundles

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun

    1994-02-01

    A knowledge of the condition leading to critical heat flux (CHF) is of great importance in the design of nuclear reactors. Although many efforts have been devoted to the subject of CHF during the last few decades, information on the burnout phenomenon at low velocity condition is very limited. Furthermore, in most cases, the applicable range of a bundle CHF correlation is restricted to a narrow region mainly due to the limitation of the CHF data base used in the correlation development. In view of these points, theoretical and experimental investigations are performed in this study for round tubes and rod bundles. A CHF prediction model for low velocity conditions is proposed throughout the assessment of CHF data from various sources with mass velocities less than 500 kg/m 2 s. The CHF data base is classified into seven groups with respect to the flow pattern characteristics at CHF conditions. CHF data for each group is analyzed by several CHF prediction models including; the flooding correlations, the flow regime transition criteria, the complete evaporation model, and the empirical correlations. At zero inlet flow or extremely low mass velocity conditions, the flooding correlation can be used for predicting CHF employing appropriate constant. In the slug or churn-turbulent flow regime, CHF seems to occur at the annular flow transition conditions. When CHF occurs at the annular flow region, the empirical correlation such as AECL CHF lookup table gives accurate predictions except for the ranges where density-wave instability is expected. A phenomenological model for the prediction of dryout locations under flooding-limited CHF condition is developed based on the liquid film dryout model and the two-phase mixture level theory. The mass and energy conservation equations are applicable to the liquid film considering no entrainment of liquid droplets from the film region. The variation of the two-phase mixture level after the onset of flooding is calculated based on

  12. In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; MacDonald, P.E.

    1980-01-01

    The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

  13. Post-test examination of the VVER-1000 fuel rod bundle CORA-W2

    International Nuclear Information System (INIS)

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.; Schanz, G.; Hagen, S.; Sepold, L.

    1995-01-01

    The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B 4 C absorber rod and the stainless steel grid spacers on the 'low-temperature' bundle damage initiation and progression. The B 4 C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occured. (orig./HP)

  14. Post-test examination of the VVER-1000 fuel rod bundle CORA-W2

    Energy Technology Data Exchange (ETDEWEB)

    Hofmann, P.; Noack, V.; Burbach, J.; Metzger, H.; Schanz, G.; Hagen, S.; Sepold, L.

    1995-08-01

    The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B{sub 4}C absorber rod and the stainless steel grid spacers on the `low-temperature` bundle damage initiation and progression. The B{sub 4}C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occured. (orig./HP)

  15. Experimental benchmark data for PWR rod bundle with spacer-grids

    International Nuclear Information System (INIS)

    Dominguez-Ontiveros, Elvis E.; Hassan, Yassin A.; Conner, Michael E.; Karoutas, Zeses

    2012-01-01

    In numerical simulations of fuel rod bundle flow fields, the unsteady Navier–Stokes equations have to be solved in order to determine the time (phase) dependent characteristics of the flow. In order to validate the simulations results, detailed comparison with experimental data must be done. Experiments investigating complex flows in rod bundles with spacer grids that have mixing devices (such as flow mixing vanes) have mostly been performed using single-point measurements. In order to obtain more details and insight on the discrepancies between experimental and numerical data as well as to obtain a global understanding of the causes of these discrepancies, comparisons of the distributions of complete phase-averaged velocity and turbulence fields for various locations near spacer-grids should be performed. The experimental technique Particle Image Velocimetry (PIV) is capable of providing such benchmark data. This paper describes an experimental database obtained using two-dimensional Time Resolved Particle Image Velocimetry (TR-PIV) measurements within a 5 × 5 PWR rod bundle with spacer-grids that have flow mixing vanes. One of the unique characteristic of this set-up is the use of the Matched Index of Refraction technique employed in this investigation to allow complete optical access to the rod bundle. This unique feature allows flow visualization and measurement within the bundle without rod obstruction. This approach also allows the use of high temporal and spatial non-intrusive dynamic measurement techniques namely TR-PIV to investigate the flow evolution below and immediately above the spacer. The experimental data presented in this paper includes explanation of the various cases tested such as test rig dimensions, measurement zones, the test equipment and the boundary conditions in order to provide appropriate data for comparison with Computational Fluid Dynamics (CFD) simulations. Turbulence parameters of the obtained data are presented in order to gain

  16. Preliminary Investigation on Turbulent Flow in Tight-lattice Rod Bundle with Twist-mixing Vane Spacer Grid

    International Nuclear Information System (INIS)

    Lee, Chiyoung; Kwack, Youngkyun; Park, Juyong; Shin, Changhwan; In, Wangkee

    2013-01-01

    Our research group has investigated the effect of P/D difference on the behavior of turbulent rod bundle flow without the mixing vane spacer grid, using PIV (Particle Image Velocimetry) and MIR (Matching Index of Refraction) techniques for tight lattice fuel rod bundle application. In this work, using the tight-lattice rod bundle with a twist-mixing vane spacer grid, the turbulent rod bundle flow is preliminarily examined to validate the PIV measurement and CFD (Computational Fluid Dynamics) simulation. The turbulent flow in the tight-lattice rod bundle with a twist-mixing vane spacer grid was preliminarily examined to validate the PIV measurement and CFD simulation. Both were in agreement with each other within a reasonable degree of accuracy. Using PIV measurement and CFD simulation tested in this work, the detailed investigations on the behavior of turbulent rod bundle flow with the twist-mixing vane spacer grid will be performed at various conditions, and reported in the near future

  17. TEGENA: Detailed experimental investigations of temperature and velocity distributions in rod bundle geometries with turbulent sodium flow

    International Nuclear Information System (INIS)

    Moeller, R.

    1989-12-01

    Precise knowlege of the velocity and temperature distributions is necessary in fuel element design (rod bundles with longitudinal flow). The detail codes required in the fine analysis of non-uniformly cooled bundle zones are presently at the stage of development. In order to verify these computer codes, the mean fluid temperatures and the related RMS values of the temperature fluctuations were measured in a heated bundle, TEGENA, containing four rods arranged in one row (P/D = W/D = 1.147) with sodium cooling (Pr≅0.005). The temperature distribution in the structures was determined as the necessary boundary condition for the temperature profiles in the fluid. The experiments were carried out with different types of heating (uniform load and flux tilting) and the flow conditions were varied in the ranges 4000≤Re≤76,000; 20≤Pe≤400. The essential processes of thermal development took place under uniform load within a heated bundle length of about 100 hydraulic diameters. In the main measuring plane at the end of the heated zone, after 200 hydraulic diameters, the flow can be termed largely developed thermally. There, the temperature profiles measured in the fluid exhibit pronounced maxima in the narrowest gaps of the subchannels as well as pronounced minima in the centers of the subchannels at the unheated wall. In the zones of maximum temperature gradients the temperature fluctuations attain maximum and minimum values, respectively, at the points of disappearance of the temperature gradients. In all cases of flux tilting investigated the flow at the end of the heated zone had not yet developed thermally. (orig.) [de

  18. Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Post test investigations of bundle test ESBU-2A

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Wallenfels, K.P.; Buescher, B.

    1986-11-01

    This KfK report describes the post test investigation of bundle experiment ESBU-2a. ESBU-2a was the second of two bundle tests on the temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of the program of out-of-pile experiments performed within the frame work of the PNS-Severe Fuel Damage program. The bundle was composed of a 3x3 fuel rod array of our fuel rod simulators (central tungsten heater, UO 2 -ring pellet and zircaloy cladding). The length was 0.4 meter. The bundle was heated to a maximum temperature of 2175 0 C. Molten cladding which dissolved part of the UO 2 pellets and slumped away from the already oxidized cladding formed a lump in the lower part of the bundle. After the test the bundle was embedded in epoxy and sectioned with a diamand saw, in the region of the refrozen melt. The cross sections were investigated by metallographic examination. The refrozen (U,Zr,O) melt consists variously of three phases with increasing oxygen content (metallic α-Zry, metallic (U,Zr) alloy and a (U,Zr)O 2 mixed oxide), two phases (α-Zry, (U,Zr)O 2 mixed oxide), or one phase ((U,Zr)O 2 mixed oxide). The cross sections show the increasing oxidation of the cladding with increasing elevation (temperature). A strong azimuthal dependency of the oxidation is found. In regions where the initial oxidized cladding is contacted by the melt one can recognize the interaction between the metallic melt and ZrO 2 of the cladding. Oxygen is taken away from the ZrO 2 . If the melt is in direct contact with steam a relatively well defined oxide layer is formed. (orig.) [de

  19. Modeling and analysis of thermal damping in heat exchanger tube bundles

    Energy Technology Data Exchange (ETDEWEB)

    Khushnood, Shahab, E-mail: seeshahab@yahoo.co [University of Engineering and Technology, Taxila (Pakistan); Khan, Zaffar Muhammad, E-mail: mafzmlk@hotmail.co [National University of Sciences and Technology, Rawalpindi (Pakistan); Malik, Muhammad Afzaal [National University of Sciences and Technology, Rawalpindi (Pakistan); Iqbal, Qamar, E-mail: qamarch@yahoo.co [University of Engineering and Technology, Taxila (Pakistan); Bashir, Sajid; Khan, Muddasar [University of Engineering and Technology, Taxila (Pakistan); Koreshi, Zafarullah, E-mail: zaffark@yahoo.co [Air University, Islamabad (Pakistan); Khan, Mahmood Anwar [National University of Sciences and Technology, Rawalpindi (Pakistan); Malik, Tahir Nadeem [University of Engineering and Technology, Taxila (Pakistan); Qureshi, Arshad Hussain [University of Engineering and Technology, Lahore (Pakistan)

    2010-07-15

    Most structures and equipment used in nuclear power plant and process plant, such as reactor internals, fuel rods, steam generator tubes bundles, and process heat exchanger tube bundles, are subjected to flow-induced vibrations (FIV). Costly plant shutdowns have been the source of motivation for continuing studies on cross-flow-induced vibration in these structures. Damping has been the target of various research attempts related to FIV in tube bundles. A recent research attempt has shown the usefulness of a phenomenon termed as 'thermal damping'. The current paper focuses on the modeling and analysis of thermal damping in tube bundles subjected to cross-flow. It is expected that the present attempt will help in establishing improved design guidelines with respect to damping in tube bundles.

  20. Experimental determination of temperature fields in sodium-cooled rod bundles with hexagonal rod arrangement and grid spacers

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.; Kolodziej, M.

    1977-01-01

    Three-dimensional temperature fields in the claddings of sodium cooled rods were determined experimentally under representative nominal operating conditions for a SNR typical 19-rod bundle model provided with spark-eroded spacers. These experiments are required to verify thermohydraulic computer programs which will provide the output data for strength calculations of the high loaded cladding tubes. In this work the essentials are reported of the measured circumferential distributions of wall temperatures of peripheral rods. In addition the sub-channel temperatures measured over the bundle cross section are indicated, they are required to sustain codes for the global thermohydraulic design of core elements. The most important results are: 1) The whole fuel element is located within the thermal entrance length. 2) High azimuthal temperature differences were measured in the claddings of peripheral rods, which are strongly influenced by the distance between the rod and the shroud, especially for the corner rod. 3) With decreasing Pe-number ( [de

  1. Behavior of a nine-rod PWR bundle under power-cooling-mismatch conditions

    International Nuclear Information System (INIS)

    Gunnerson, F.S.; Sparks, D.T.

    1979-01-01

    An experiment to characterize the behavior of a nine-rod pressurized water reactor (PWR) fuel bundle operating during power-cooling-mismatch (PCM) conditions has been conducted in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The experiment, designated Test PCM-5, is part of a series of PCM experiments designed to evaluate light water reactor (LWR) fuel rod response under postulated accident conditions. Test PCM-5 was the first nine-rod bundle experiment in the PCM test series. The primary objectives and the results of the experiment are described

  2. Posttest examination of the VVER-1000 fuel rod bundle CORA-W2

    International Nuclear Information System (INIS)

    Sepold, L.

    1995-06-01

    The bundle meltdown experiment CORA-W2, representing the behavior of a Russian type VVER-1000 fuel element, with one B 4 C/stainless steel absorber rod was selected by the OECD/CSNI as International Standard Problem (ISP-36). The experimental results of CORA-W2 serve as data base for comparison with analytical predictions of the high-temperature material behavior by various code systems. The first part of the experimental results is described in KfK 5363 (1994), the second part is documented in this report which contains the destructive post-test examination results. The metallographical and analytical (SEM/EDX) post-test examinations were performed in Germany and Russia and are summarized in five individual contributions. The upper half of the bundle is completely oxidized, the lower half has kept the fuel rods relatively intact. The post-test examination results show the strong impact of the B 4 C absorber rod and the stainless steel grid spacers on the ''low-temperature'' bundle damage initiation and progression. The B 4 C absorber rod completely disappeared in the upper half of the bundle. The multicomponent melts relocated and formed coolant channel blockages on solidification with a maximum extent of about 30% in the lower part of the bundle. At temperatures above the melting point of the ZrNb1 cladding extensive fuel dissolution occurred. (orig.) [de

  3. TEGENA: Detailed experimental investigations of temperature and velocity distributions in rod bundle geometries with turbulent sodium flow

    International Nuclear Information System (INIS)

    Moeller, R.

    1989-02-01

    Precise knowledge of the velocity and temperature distributions is necessary in fuel element design (rod bundles with longitudinal flow). The detail codes required in the fine analysis of non-uniformly cooled bundle zones are presently at the stage of development. In order to verify these computer codes, the mean fluid temperatures and the related RMS values of the temperature fluctuations were measured in a heated bundle TEGENA, containing 4 rods arranged in one row (P/D = W/D = 1.147) with sodium cooling (Pr ≅ 0.005). The temperature distribution in the structures was determined as the necessary boundary condition for the temperature profiles in the fluid. The experiments were carried out with different types of heating (uniform load and load tilting) and the flow conditions were varied in the range from 4000 ≤ Re ≤ 76.000, 20 ≤ Pe ≤ 400. The essential process of thermal development took place under uniform load within a heated bundle length of about 100 hydraulic diameters. In the main measuring plane at the end of the heated zone, after 200 hydraulic diameters, the flow can be termed largely developed thermally. There, the temperature profiles measured in the fluid exhibit pronounced maxima in the narrowest gaps of the subchannels as well as pronounced minima in the centers of the subchannels at the unheated wall. In the zones of maximum temperature gradients the temperature fluctuations attain maximum and minimum values, respectively, at the points of disappearance of the temperature gradients. In all cases of load tilting investigated the flow at the end of the heated zone had not yet developed thermally. By inspection of all thermocouples in isothermal experiments performed at regular intervals, by redundant arrangement of the mobile probe thermocouples and by demonstration of the reproducibility of results of measurement the experiments have been validated satisfactorily. (orig./GL) [de

  4. Prediction of interfacial area transport in a scaled 8×8 BWR rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Yang, X.; Schlegel, J.P.; Liu, Y.; Paranjape, S.; Hibiki, T.; Ishii, M. [School of Nuclear Engineering, Purdue University, 400 Central Dr., West Lafayette, IN 47907-2017 (United States); Bajorek, S.; Ireland, A. [U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 (United States)

    2016-12-15

    In the two-fluid model, it is important to give an accurate prediction for the interfacial area concentration. In order to achieve this goal, the interfacial area transport equation has been developed. This study focuses on the benchmark of IATE performance in a rod bundle geometry. A set of interfacial area concentration source and sink term models are proposed for a rod bundle geometry based on the confined channel IATE model. This model was selected as a basis because of the relative similarity of the two geometries. Benchmarking of the new model with interfacial area concentration data in an 8×8 rod bundle test section which has been scaled from an actual BWR fuel bundle is performed. The model shows good agreement in bubbly and cap-bubbly flows, which are similar in many types of geometries, while it shows some discrepancy in churn-turbulent flow regime. This discrepancy may be due to the geometrical differences between the actual rod bundle test facility and the facility used to collect the data which benchmarked the original source and sink models.

  5. Correlations of drift velocity for gas-liquid two-phase flow in rod bundle

    International Nuclear Information System (INIS)

    Kataoka, Isao; Matsuura, Keizo; Serizawa, Akimi

    2004-01-01

    A new correlation was developed for the drift velocity for low inlet liquid flux in rod bundle. Based on authors' previous analysis of drift velocity for large diameter pipe, an analysis was made on the drift velocity in rod bundle. It is assumed that the large bubble of which size is several subchannel diameter behaves as slug bubble. Under this assumption, it becomes very important how to define equivalent diameter for rod bundle. In view of physical consideration of slug bubble behavior and previous analysis, an equivalent diameter based on the wetted perimeter was found to be most appropriate. Using this equivalent diameter, experimental data of drift velocity in rod bundle were correlated with dimensional analysis. It was found out that for small diameter (dimensionless diameter less than 48) drift velocity increased with square root of diameter which is same dependency of ordinary slug flow correlation. For larger diameter (dimensionless diameter is more than 48), drift velocity is almost constant and same as that of dimensionless diameter of 48. The physical meaning of this result was considered to be the instability of interface of large slug bubble. The density ratio between gas and liquid and viscosity of liquid phase were found to be the main parameters which affect the drift velocity. This is physically reasonable because density ratio is related to the buoyancy force and liquid viscosity is related to shear force near solid wall. The experimental data were correlated by density ratio and dimensionless liquid viscosity. The obtained dimensionless correlation for the drift velocity in rod bundle successfully correlated experimental data for various rod bundles (equivalent diameters), pressures, liquid fluxes etc. It is also consistent with the drift flux correlation for round tube. (author)

  6. Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Test ESBU-1

    International Nuclear Information System (INIS)

    Hagen, S.; Malauschek, H.; Peck, S.O.; Wallenfels, K.P.

    1983-12-01

    This report describes the test conduct and results of the bundle test ESBU-1. The test objective was the investigation of temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of a program of out-of-pile experiments, performed within the framework of the PNS Several Fuel Damage Program. The bundle was composed of a 3x3 array of fuel rod simulators surrounded by a zircaloy shroud which was insulated with a ZrO 2 fiber ceramic wrap. The fuel rod simulators comprised a tungsten heater, UO 2 annular pellets, and zircaloy cladding over a 0.4 m heated length. A steam flow of 1 g/s was inlet to the bundle. The most pronounced temperature escalation was found on the central rod. The initial heatup rate of 2 0 C/s at 1100 0 C increased to approximately 6 0 C/s. The maximum temperature reached was 2250 0 C. The following fast temperature decrease was caused by runoff of molten zircaloy. Molten zircaloy swept down the thin cladding oxide layer formed during heatup. The melt dissolved the surface of the UO 2 pellets and refroze as a coherent lump in the lower part of the bundle. The remaining pellets fragmented during cooldown and formed a powdery layer on the refrozen lump. The lump was sectioned posttest at several elevations: Dissolution of UO 2 by the molten zircaloy, interaction between the melt and previously oxidized zircaloy, and oxidation of the melt had occurred. (orig.) [de

  7. Hydrodynamics around a spacer of a VVER-440 fuel rod bundle

    International Nuclear Information System (INIS)

    Mayer, G.; Hazi, G.; Kavran, P.

    2004-01-01

    Recently, an intensive research has been started in our institute, focusing on the hydrodynamics of fuel rod bundles. Numerical computations have been planed to be carried out in a three level bottom-up hierarchy, using direct numerical simulation, large eddy simulation and Reynolds averaged Navier-Stokes approach. Here, we give a description of the numerical method applied for direct numerical and large eddy simulation. We present some preliminary results obtained by the simulation of the flow around a spacer of a VVER-440 fuel rod bundle. (author)

  8. Heat transfer on HLM cooled wire-spaced fuel pin bundle simulator in the NACIE-UP facility

    Energy Technology Data Exchange (ETDEWEB)

    Di Piazza, Ivan, E-mail: ivan.dipiazza@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone, Camugnano (Italy); Angelucci, Morena; Marinari, Ranieri [University of Pisa, Dipartimento di Ingegneria Civile e Industriale, Pisa (Italy); Tarantino, Mariano [Italian National Agency for New Technologies, Energy and Sustainable Economic Development, C.R. ENEA Brasimone, Camugnano (Italy); Forgione, Nicola [University of Pisa, Dipartimento di Ingegneria Civile e Industriale, Pisa (Italy)

    2016-04-15

    Highlights: • Experiments with a wire-wrapped 19-pin fuel bundle cooled by LBE. • Wall and bulk temperature measurements at three axial positions. • Heat transfer and error analysis in the range of low mass flow rates and Péclet number. • Comparison of local and section-averaged Nusselt number with correlations. - Abstract: The NACIE-UP experimental facility at the ENEA Brasimone Research Centre (Italy) allowed to evaluate the heat transfer coefficient of a wire-spaced fuel bundle cooled by lead-bismuth eutectic (LBE). Lead or lead-bismuth eutectic are very attractive as coolants for the GEN-IV fast reactors due to the good thermo-physical properties and the capability to fulfil the GEN-IV goals. Nevertheless, few experimental data on heat transfer with heavy liquid metals (HLM) are available in literature. Furthermore, just a few data can be identified on the specific topic of wire-spaced fuel bundle cooled by HLM. Additional analysis on thermo-fluid dynamic behaviour of the HLM inside the subchannels of a rod bundle is necessary to support the design and safety assessment of GEN. IV/ADS reactors. In this context, a wire-spaced 19-pin fuel bundle was installed inside the NACIE-UP facility. The pin bundle is equipped with 67 thermocouples to monitor temperatures and analyse the heat transfer behaviour in different sub-channels and axial positions. The experimental campaign was part of the SEARCH FP7 EU project to support the development of the MYRRHA irradiation facility (SCK-CEN). Natural and mixed circulation flow regimes were investigated, with subchannel Reynolds number in the range Re = 1000–10,000 and heat flux in the range q″ = 50–500 kW/m{sup 2}. Local Nusselt numbers were calculated for five sub-channels in different ranks at three axial positions. Section-averaged Nusselt number was also defined and calculated. Local Nusselt data showed good consistency with some of the correlation existing in literature for heat transfer in liquid metals

  9. Heat transfer on HLM cooled wire-spaced fuel pin bundle simulator in the NACIE-UP facility

    International Nuclear Information System (INIS)

    Di Piazza, Ivan; Angelucci, Morena; Marinari, Ranieri; Tarantino, Mariano; Forgione, Nicola

    2016-01-01

    Highlights: • Experiments with a wire-wrapped 19-pin fuel bundle cooled by LBE. • Wall and bulk temperature measurements at three axial positions. • Heat transfer and error analysis in the range of low mass flow rates and Péclet number. • Comparison of local and section-averaged Nusselt number with correlations. - Abstract: The NACIE-UP experimental facility at the ENEA Brasimone Research Centre (Italy) allowed to evaluate the heat transfer coefficient of a wire-spaced fuel bundle cooled by lead-bismuth eutectic (LBE). Lead or lead-bismuth eutectic are very attractive as coolants for the GEN-IV fast reactors due to the good thermo-physical properties and the capability to fulfil the GEN-IV goals. Nevertheless, few experimental data on heat transfer with heavy liquid metals (HLM) are available in literature. Furthermore, just a few data can be identified on the specific topic of wire-spaced fuel bundle cooled by HLM. Additional analysis on thermo-fluid dynamic behaviour of the HLM inside the subchannels of a rod bundle is necessary to support the design and safety assessment of GEN. IV/ADS reactors. In this context, a wire-spaced 19-pin fuel bundle was installed inside the NACIE-UP facility. The pin bundle is equipped with 67 thermocouples to monitor temperatures and analyse the heat transfer behaviour in different sub-channels and axial positions. The experimental campaign was part of the SEARCH FP7 EU project to support the development of the MYRRHA irradiation facility (SCK-CEN). Natural and mixed circulation flow regimes were investigated, with subchannel Reynolds number in the range Re = 1000–10,000 and heat flux in the range q″ = 50–500 kW/m"2. Local Nusselt numbers were calculated for five sub-channels in different ranks at three axial positions. Section-averaged Nusselt number was also defined and calculated. Local Nusselt data showed good consistency with some of the correlation existing in literature for heat transfer in liquid metals for

  10. Full-scale model development of the WWER-440 reactor fuel rod bundle for core temperature regime study under reflooding conditions

    International Nuclear Information System (INIS)

    Bezrukov, Yu.A.; Logvinov, S.A.; Levchuk, S.V.; Nakladnov, V.D.; Onshin, V.P.; Sokolov, A.S.

    1982-01-01

    Consideration is given to the issues of a full scale WWER-440 fuel rod bundle imitation. An imitator contains a molybdenum heating rod inclosed in stainless steel shell. The shell diameter is 9 mm, the heated length is 2500 mm, the total len.o.th is 2855 mm. 125 fuel rod imitators are set in the bundle mock-up. The experiments were run on a test facility imitating the WWER-440 reactor primary loop, providing the conditions of the loop breaking. The mock-up thermal hydraulics has been studied during the refloodino. stage. The mock-up was heated up to predetermined initial temperature at a low power level with saturated steam cooling. Then the steam input was stopped, the power level rarapidly rised up to a given value and the cooling water injected. Simultaneously with water injection all the measured parameters monitoring was started. Both at the top spraying and combined cooling temperature oscillations in the upper and middle parts of the mock-up were observed. At the bottom reflooding the mock-up cooling down took more time, thereat temperature inthe upper part first slowly rised during reflooding then decreased and then dropped abruptly at thefront coming up [ru

  11. Computer code TOBUNRAD for PWR fuel bundle heat-up calculations

    International Nuclear Information System (INIS)

    Shimooke, Takanori; Yoshida, Kazuo

    1979-05-01

    The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)

  12. Experiment and numerical simulation of bubbly two-phase flow across horizontal and inclined rod bundles

    International Nuclear Information System (INIS)

    Serizawa, A.; Huda, K.; Yamada, Y.; Kataoka, I.

    1997-01-01

    Experimental and numerical analyses were carried out on vertically upward air-water bubbly two-phase flow behavior in both horizontal and inclined rod bundles with either in-line or staggered array. The inclination angle of the rod bundle varied from 0 to 60 with respect to the horizontal. The measured phase distribution indicated non-uniform characteristics, particularly in the direction of the rod axis when the rods were inclined. The mechanisms for this non-uniform phase distribution is supposed to be due to: (1) Bubble segregation phenomenon which depends on the bubble size and shape: (2) bubble entrainment by the large scale secondary flow induced by the pressure gradient in the horizontal direction which crosses the rod bundle; (3) effects of bubble entrapment by vortices generated in the wake behind the rods which travel upward along the rod axis; and (4) effect of bubble entrainment by local flows sliding up along the front surface of the rods. The liquid velocity and turbulence distributions were also measured and discussed. In these speculations, the mechanisms for bubble bouncing at the curved rod surface and turbulence production induced by a bubble were discussed, based on visual observations. Finally, the bubble behaviors in vertically upward bubbly two-phase flow across horizontal rod bundle were analyzed based on a particle tracking method (one-way coupling). The predicted bubble trajectories clearly indicated the bubble entrapment by vortices in the wake region. (orig.)

  13. CFD simulation of flow and heat transfer in Canadian SCWR bundles

    International Nuclear Information System (INIS)

    Podila, K.; Rao, Y.F.

    2014-01-01

    Within the Generation-IV (Gen-IV) International Forum, Atomic Energy of Canada Limited (AECL) is leading the effort in developing a conceptual design for the Canadian supercritical water-cooled reactor (SCWR). AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for possible occurrence of heat transfer deterioration (HTD) in the supercritical bundle flows. In the current investigation, a bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The SST k-ω turbulence model along with y + <1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5MPa). Ultimately, the goal of this study is to aid the development of a criterion for the onset of HTD in the proposed SCWR bundles, which is planned in the next phase of the project. (author)

  14. Critical power characteristics in 37-rod tight lattice bundles under transient conditions

    International Nuclear Information System (INIS)

    Liu, Wei; Kureta, Masatoshi; Tamai, Hidesada; Ohnuki, Akira; Akimoto, Hajime

    2007-01-01

    Critical power characteristics in the postulated abnormal transient processes that may be possibly met in the operation of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) were investigated for the design of the FLWR core. Transient Boiling Transition (BT) tests were carried out using two sets of 37-rod tight lattice rod bundles (rod diameter: 13 mm; rod clearance: 1.3 mm or 1.0 mm) at Japan Atomic Energy Agency (JAEA) under the conditions covering the FLWR operating condition (P ex =7.2 MPa, T in =556 K) for mass velocity G=400-800 kg/(m 2 s). For the postulated power increase and flow decrease transients, no obvious change of the critical power against the steady one was observed. The traditional quasi-steady characteristic was confirmed to be working for the postulated power increase and flow decrease transients. The experiments were analyzed with TRAC-BF1 code, where the JAEA newest critical power correlation for the tight lattice rod bundles was implemented for the BT judgment. The TRAC-BF1 code showed good prediction for the occurrence or the non occurrence of the BT and for the exact BT starting time. The tranditional quasi-steady state prediction of the BT in transient process was confirmed to be applicable for the postulated abnormal transient processes in the tight lattice rod bundles. (author)

  15. ORNL rod-bundle heat-transfer test data. Volume 7. Thermal-Hydraulic Test Facility experimental data report for test series 3.07.9 - steady-state film boiling in upflow

    International Nuclear Information System (INIS)

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

    1982-05-01

    Thermal-Hydraulic Test Facility (THTF) test series 3.07.9 was conducted by members of the Oak Ridge National Laboratory Pressurized-Water Reactor (ORNL-PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on September 11, September 18, and October 1, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small- and large-break loss-of-coolant accidents. Test series 3.07.9 was designed to provide steady-state film boiling data in rod bundle geometry under reactor accident-type conditions. This report presents the reduced instrument responses for THTF test series 3.07.9. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

  16. ORNL rod-bundle heat-transfer test data. Volume 3. Thermal-hydraulic test facility experimental data report for test 3.06.6B - transient film boiling in upflow

    International Nuclear Information System (INIS)

    Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

    1982-05-01

    Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

  17. Pressure drop ana velocity measurements in KMRR fuel rod bundles

    International Nuclear Information System (INIS)

    Yagn, Sun Kyu; Chung, Heung June; Chung, Chang Whan; Chun, Se Young; Song, Chul Wha; Won, Soon Yeun; Chung, Moon Ki

    1990-01-01

    The detailed hydraulic characteristic measurements in subchannels of longitudinally finned rod bundles using one-component LDV(Laser Doppler Velocimeter) were performed. Time mean axial velocity, turbulent intensity, and turbulent micro scales, such as time auto-correlation, Eulerian integral and micro scale, Kolmogorov length and time scale, and Taylor micro length scale were measured. The signals from LDV are inherently more or less discontinuous. The spectra of signals having such intermittent defects can be obtained by the fast Fourier transformation (FFT) of the auto-correlation function. The turbulent crossflow mixing rate between neighboring subchannels and dominant frequencies were evaluated from the measured data. Pressure drop data were obtained for the typical 36-element and 18-element fuel rod bundles fabricated by the design requirement of KMRR fuel and for other type of fuels assembled with 6-fin rods to investigate the fin effects on the pressure drop characteristics

  18. Development and Assessment of a Bundle Correction Method for CHF

    International Nuclear Information System (INIS)

    Hwang, Dae Hyun; Chang, Soon Heung

    1993-01-01

    A bundle correction method, based on the conservation laws of mass, energy, and momentum in an open subchannel, is proposed for the prediction of the critical heat flux (CHF) in rod bundles from round tube CHF correlations without detailed subchannel analysis. It takes into account the effects of the enthalpy and mass velocity distributions at subchannel level using the first dericatives of CHF with respect to the independent parameters. Three different CHF correlations for tubes (Groeneveld's CHF table, Katto correlation, and Biasi correlation) have been examined with uniformly heated bundle CHF data collected from various sources. A limited number of GHE data from a non-uniformly heated rod bundle are also evaluated with the aid of Tong's F-factor. The proposed method shows satisfactory CHF predictions for rod bundles both uniform and non-uniform power distributions. (Author)

  19. Characteristics of liquid and boiling sodium flows in heating pin bundles

    International Nuclear Information System (INIS)

    Menant, Bernard

    1976-01-01

    This study is related to cooling accidents which could occur in sodium cooled fast reactors. Thermo-hydraulic aspects of boiling experiments in pin bundles with helical wire-wrap spacer systems, in the case of undamaged geometries, are analyzed. Differences and analogies in the behavior of multi-rod bundle flows and one-dimensional channel flows are studied. A boiling model is developed for bundle geometries, and predictions obtained with the FLICA code using this models are presented. These predictions are compared with experimental results obtained in a water 19-rod bundle. Then, results of sodium boiling experiments through a 19-rod bundle are interpreted. Both cases of high power and reduced power are envisaged. (author) [fr

  20. Lateral Flow Field Behavior Downstream of Mixing Vanes In a Simulated Nuclear Fuel Rod Bundle

    International Nuclear Information System (INIS)

    Conner, Michael E.; Smith, L. David III; Holloway, Mary V.; Beasley, Donald E.

    2004-01-01

    To assess the fuel assembly performance of PWR nuclear fuel assemblies, average subchannel flow values are used in design analyses. However, for this highly complex flow, it is known that local conditions around fuel rods vary dependent upon the location of the fuel rod in the fuel assembly and upon the support grid design that maintains the fuel rod pitch. To investigate the local flow in a simulated nuclear fuel rod bundle, a testing technique has been employed to measure the lateral flow field in a 5 x 5 rod bundle. Particle Image Velocimetry was used to measure the lateral flow field downstream of a support grid with mixing vanes for four unique subchannels in the 5 x 5 bundle. The dominant lateral flow structures for each subchannel are compared in this paper including the decay of these flow structures. (authors)

  1. Assessment of 4x4 rod bundle subchannel mixing experiments

    International Nuclear Information System (INIS)

    Otero, Fatima; Veloso, Maria A.; Pereira, Claubia; Fortini, Angela; Lombardi, Antonella

    2011-01-01

    An assessment of mixing data taking from a 4x4 rod bundle array, under operating conditions typical of a Boiling Water Reactor (BWR), conducted at Columbia University Heat Transfer Research Facility has been accomplished by using the STHIRP-1 code, which is a UFMG version of the COBRA-3C subchannel code. Although designed for subchannel analysis of research reactor cores, all the capability of COBRA-3C has been preserved in the STHIRP-1 code. In the light of alternative models for turbulent mixing, steam quality, and void fraction, results predicted by this code will be compared with experimental data for specific enthalpy and mass flow rate measured at the exit of two specific subchannels.(author)

  2. Numerical Analysis for IFM Grid Effect on 5x5 Rods Bundle

    International Nuclear Information System (INIS)

    Kim, Seong Jin; Cha, Jeong Hun; Seo, Kyong Won; Kim, Tae Woo; Kwon, Hyuk; Hwang, Dae Hyun

    2011-01-01

    Generally, the fuel assembly consists of fuel rods, bottom and top grids, spacer grids, mixing vane, etc. The mixing vane with spacer grid is used to increase the thermal mixing between subchannels and to increase CHF(Critical Heat Flux). IFM(Intermediate Flow Mixer) grids are used to induce lateral flow between adjacent channels and are well-known as improving CHF, also. A numerical analysis using CFD code(ANSYS CFX, version 12.1) and subchannel code(MATRA-S) was conducted to investigate the influence of IFM grid on the subchannel temperature in 5x5 rods bundle with and without the IFM grid, thermohydraulically. In this study, the quantitative improvement of the mixing effect of the IFM grid is presented from the results of CFX and MATRA-S code. Moreover, capacity of predicting subchannel temperature of MATRA-S code is compared with CFX result

  3. Thyc, a 3D thermal-hydraulic code for rod bundles. Recent developments and validation tests

    International Nuclear Information System (INIS)

    Caremoli, C.; Rascle, P.; Aubry, S.; Olive, J.

    1993-09-01

    PWR or LMFBR cores or fuel assemblies, PWR steam generators, condensers, tubular heat exchangers, are basic components of a nuclear power plant involving two-phase flows in tube or rod bundles. A deep knowledge of the detailed flow patterns on the shell side is necessary to evaluate DNB margins in reactor cores, singularity effects (grids, wire spacers, support plates, baffles), corrosion on steam generator tube sheet, bypass effects and vibration risks. For that purpose, Electricite de France has developed, since 1986, a general purpose code named THYC (Thermal HYdraulic Code) designed to study three-dimensional single and two phase flows in rod or tube bundles (pressurized water reactor cores, steam generators, condensers, heat exchangers). It considers the three-dimensional domain to contain two kinds of components: fluid and solids. The THYC model is obtained by space-time averaging of the instantaneous equations (mass, momentum and energy) of each phase over control volumes including fluid and solids. This paper briefly presents the physical model and the numerical method used in THYC. Then, validation tests (comparison with experiments) and applications (coupling with three-dimensional neutronics code and DNB predictions) are presented. They emphasize the last developments and new capabilities of the code. (authors). 10 figs., 3 tabs., 21 refs

  4. Air-water two-phase flow in a four by four rod bundle with partial length rods

    International Nuclear Information System (INIS)

    Ohta, Motoki; Kamei, Akihiro; Mizutani, Yoshitaka; Hosokawa, Shigeo; Tomiyama, Akio

    2009-01-01

    Partial length rods (PLR) are used in fuel bundles of BWR to reduce pressure drops in two-phase regions and to optimize the power distribution. Since little is known about effects of PLR on two-phase flows, air-water two-phase flow around PLRs in a four by four rod bundle is visualized by using a high-speed video camera. The experimental apparatus consists of acrylic channel box and transparent rods. Air and water at atmospheric pressure and room temperature are used for the gas and liquid phases, respectively. The ranges of the gas and liquid volume fluxes, J G and J L , are 0.4 L G L , the flow pattern in the downstream of PLR transits to slug flow, and the flow patterns in the surrounding subchannels transit to bubbly flow due to the redistribution of gas flow. (2) In annular flow, the liquid film on the PLR forms a liquid column above the end cap of PLR. Droplets are generated by column breakup and deposit on liquid films on the neighboring rods. (3) The liquid film thickness on the surface of neighbor rods facing the PLR increases and it reduces that on their opposite surface in the downstream of PLR. (author)

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

    International Nuclear Information System (INIS)

    Shen Xiuzhong; Yu Pingan; Yang Guanyue

    1996-01-01

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

  6. Boiling heat transfer on horizontal tube bundles

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    Nucleate boiling heat transfer characteristics for a tube in a bundle differ from that for a single tube in a pool and this difference is known as 'tube bundle effect.' There exist two bundle effects, positive and negative. The positive bundle effect enhances heat transfer due to convective flow induced by rising bubbles generated from the lower tubes, while the negative bundle effect deteriorates heat transfer due to vapor blanketing caused by accumulation of bubbles. Staggered tube bundles tested and found that the upper tubes in bundles have higher heat transfer coefficients than the lower tubes. The effects of various parameters such as pressure, tube geometry and oil contamination on heat transfer have been examined. Some workers attempted to clarify the mechanism of occurrence of 'bundle effect' by testing tube arrangements of small scale. All reported only enhancement in heat transfer but results showed the symptom of heat transfer deterioration at higher heat fluxes. As mentioned above, it has not been clarified so far even whether the 'tube bundle effect' should serve as enhancement or deterioration of heat transfer in nucleate boiling. In this study, experiments are performed in detail by using bundles of small scale, and effects of heat flux distribution, pressure and tube location are clarified. Furthermore, some consideration on the mechanisms of occurrence of 'tube bundle effect' is made and a method for prediction of heat transfer rate is proposed

  7. DESIGN OF WIRE-WRAPPED ROD BUNDLE MATCHED INDEX-OF-REFRACTION EXPERIMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Hugh McIlroy; Hongbin Zhang; Kurt Hamman

    2008-05-01

    Experiments will be conducted in the Idaho National Laboratory (INL) Matched Index-of-Refraction (MIR) Flow Facility [1] to characterize the three-dimensional velocity and turbulence fields in a wire-wrapped rod bundle typically employed in liquid-metal cooled fast reactors and to provide benchmark data for computer code validation. Sodium cooled fast reactors are under consideration for use in the U.S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP) program. The experiment model will be constructed of quartz components and the working fluid will be mineral oil. Accurate temperature control (to within 0.05 oC) matches the index-of-refraction of mineral oil with that of quartz and renders the model transparent to the wavelength of laser light employed for optical measurements. The model will be a scaled 7-pin rod bundle enclosed in a hexagonal canister. Flow field measurements will be obtained with a LaVision 3-D particle image velocimeter (PIV) and complimented by near-wall velocity measurements obtained from a 2-D laser Doppler velocimeter (LDV). These measurements will be used as benchmark data for computational fluid dynamics (CFD) validation. The rod bundle model dimensions will be scaled up from the typical dimensions of a fast reactor fuel assembly to provide the maximum Reynolds number achievable in the MIR flow loop. A range of flows from laminar to fully-turbulent will be available with a maximum Reynolds number, based on bundle hydraulic diameter, of approximately 22,000. The fuel pins will be simulated by 85 mm diameter quartz tubes (closed on the inlet ends) and the wire-wrap will be simulated by 25 mm diameter quartz rods. The canister walls will be constructed from quartz plates. The model will be approximately 2.13 m in length. Bundle pressure losses will also be measured and the data recorded for code comparisons. The experiment design and preliminary CFD calculations, which will be used to provide qualitative hydrodynamic

  8. Experimental investigations of turbulent flows in rod bundles with and without spacer grids

    International Nuclear Information System (INIS)

    Trippe, G.

    1979-07-01

    In the thermofluiddynamic design of liquid metal cooled reactor fuel elements the lack of experimentally confirmed knowledge of the three-dimensional flow events in rod bundles provided with spacer grids has appeared as a significant problem. To close this gap of knowledge, detailed measurements of the local velocities were made on a 19-rod bundle model. The Pitot method of differential pressure measurements was used as the measuring system. In the first part of the work the fully developed flow regime not influenced by spacers was investigated. A simple relation was derived for distributing the mass flow among the subchannels of a rod bundle; it is but slightly dependent on the Reynolds number. This relation allows a quick, coarse calculation of the distribution of the undisturbed, fully developed mass flow in bundles with similar geometries. By evaluation of further experiments known from the literature, empirical relationships were found for the local velocity distribution within the subchannels of such bundles. In the second part the effect of grid shaped spacers was investigated. The three-dimensional flow events caused by the spacers were completely recorded and interpreted physically. The deeper understanding of these flow processes can now serve to improve the model concept used in the present design computer programs. Single results of the investigations which take primary importance are the quantitative relations existing between the changes of mass flow in the bundle boundary zone, caused by a spacer, and the geometry of this spacer. The transferability to other bundle geometries was discussed and delimited. Moreover, it was shown that the mass flow in the bundle boundary zone can be successively reduced by spacers placed one behind the other in the bundle. A noticeable dependence of flow events on the Reynolds number was not found for the range relevant in practical application (30.000 [de

  9. Laboratory manual for salt mixing test in rod bundles

    International Nuclear Information System (INIS)

    Khan, H.U.R.; Chiu, C.; Todreas, N.

    1978-10-01

    This report is a Laboratory Manual dealing with the procedure employed during Salt Tracer Experiments, which are used for evaluating the hydraulic characteristics of a rod bundle. A description of the standard equipment used is given together with details of manufacture of non-standard items i.e., probes used for detecting the salt-concentration. Details of bundle construction have not been included as they are available in the references cited. An attempt has also been made to point out potential trouble areas and procedures

  10. Experimental investigation of turbulent flow through spacer grids in fuel rod bundles

    International Nuclear Information System (INIS)

    Caraghiaur, Diana; Anglart, Henryk; Frid, Wiktor

    2009-01-01

    This paper contains experimental data of pressure, velocity and turbulence intensity in a 24-rod fuel bundle with spacer grids. Detailed pressure measurements inside the spacer grid have been obtained by use of a sliding pressure-sensing rod. Laser Doppler Velocimetry technique was used to measure the local axial velocity and its fluctuating component upstream and downstream of the spacer grid in sub-channels with different blockage ratios. The measurements show a changing pattern in function of radial position in the cross-section of the fuel bundle. For sub-channels close to the box wall, the turbulence intensity suddenly increases just downstream of the spacer and then gradually decays. In inner sub-channels, however, the turbulence intensity downstream of the spacer decreases below its upstream value and then gradually increases until it reaches the maximum value at approximately two spacer heights. The present study reveals that spacer effects, such as local pressure distribution and turbulence intensity enhancement, not only depend exclusively on the local geometry details, but also on the location in the cross-section of the rod bundle.

  11. Experimental investigation of turbulent flow through spacer grids in fuel rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Caraghiaur, Diana [Royal Institute of Technology, Division of Nuclear Reactor Technology, Department of Physics, School of Engineering Sciences, AlbaNova University Center, SE-106 91 Stockholm (Sweden)], E-mail: dianac@kth.se; Anglart, Henryk [Royal Institute of Technology, Division of Nuclear Reactor Technology, Department of Physics, School of Engineering Sciences, AlbaNova University Center, SE-106 91 Stockholm (Sweden); Frid, Wiktor [Swedish Radiation Safety Authority, Reactor Technology and Structural Integrity, SE-171 16 Stockholm (Sweden)

    2009-10-15

    This paper contains experimental data of pressure, velocity and turbulence intensity in a 24-rod fuel bundle with spacer grids. Detailed pressure measurements inside the spacer grid have been obtained by use of a sliding pressure-sensing rod. Laser Doppler Velocimetry technique was used to measure the local axial velocity and its fluctuating component upstream and downstream of the spacer grid in sub-channels with different blockage ratios. The measurements show a changing pattern in function of radial position in the cross-section of the fuel bundle. For sub-channels close to the box wall, the turbulence intensity suddenly increases just downstream of the spacer and then gradually decays. In inner sub-channels, however, the turbulence intensity downstream of the spacer decreases below its upstream value and then gradually increases until it reaches the maximum value at approximately two spacer heights. The present study reveals that spacer effects, such as local pressure distribution and turbulence intensity enhancement, not only depend exclusively on the local geometry details, but also on the location in the cross-section of the rod bundle.

  12. Two-phase flow patterns in a four by four rod bundle

    International Nuclear Information System (INIS)

    Mizutani, Yoshitaka; Tomiyama, Akio; Hosokawa, Shigeo; Sou, Akira; Kudo, Yoshiro; Mishima, Kaichiro

    2007-01-01

    Air-water two-phase flow patterns in a four by four square lattice rod bundle consisting of an acrylic channel box of 68 mm in width and transparent rods of 12mm in diameter were observed by utilizing a high speed video camera, FEP (fluorinated ethylene propylene) tubes for rods, and a fiberscope inserted in a rod. The FEP possesses the same refractive index as water, and thereby, whole flow patterns in the bundle and local flow patterns in subchannels were successfully visualized with little optical distortion. The ranges of gas and liquid volume fluxes, (J G ) and (J L ), in the present experiments were 0.1 L ) G ) G )-(J L ) flow pattern diagram is so narrow that it can be regarded as a boundary between bubbly and churn flows. (2) the boundary between bubbly and churn flows is close to the boundary between bubbly and slug flows of the Mishima and Ishii's flow pattern transition model, and (3) the boundary between churn and annular flow is close to the Mishima and Ishii's model. (author)

  13. Two-Phase Flow Patterns in a Four by Four Rod Bundle

    International Nuclear Information System (INIS)

    Yoshitaka Mizutani; Shigeo Hosokawa; Akio Tomiyama

    2006-01-01

    Air-water two-phase flow patterns in a four by four square lattice rod bundle consisting of an acrylic channel box of 68 mm in width and transparent rods of 12 mm in diameter were observed by utilizing a high speed video camera, FEP (fluorinated ethylene propylene) tubes for rods, and a fiber-scope inserted in a rod. The FEP possesses the same refractive index as water, and thereby, whole flow patterns in the bundle and local flow patterns in subchannels were successfully visualized with little optical distortion. The ranges of liquid and gas volume fluxes, G > and L >, in the present experiments were 0.1 L > G > G > - L > flow pattern diagram is so narrow that it can be regarded as a boundary between bubbly and churn flows, (2) the boundary between bubbly and churn flows is close to the boundary between bubbly and slug flows of the Mishima and Ishii's flow pattern transition model, and (3) the boundary between churn and annular flows is well predicted by the Mishima and Ishii's model. (authors)

  14. Development of design technology on thermal-hydraulic performance in tight-lattice rod bundle. III - Numerical estimation on rod bowing effect based on X-ray CT data

    International Nuclear Information System (INIS)

    Misawa, Takeharu; Ohnuki, Akira; Katsuyama, Kozo; Nagamine, Tsuyoshi; Nakamura, Yasuo; Akimoto, Hajime; Mitsutake, Toru; Misawa, Susumu

    2007-01-01

    Design studies of the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) are being carried out at the Japan Atomic Energy Agency (JAEA) as one candidate for the future reactors. In actual core design, it is precondition to prevent fuel rods contact due to fuel rod bowing. However, the FLWR cores have nonconventional characteristics such as a hexagonal tight lattice arrangement and a high enrichment fuel loading. Therefore, as conservative evaluation, it is important to investigate influence of fuel rod bowing upon the boiling transition. In the JAEA, a 37-rod bundle experiments (base case test section (1.3mm gap width), gap width effect test section (1.0mm gap width), and rod bowing test section) were performed in order to investigate the thermal hydraulic characteristics in the tight lattice bundle. In this paper, the rod bowing effect test is paid attention. It is suspected that the actual fuel rod positions in the rod bowing test section may be different from the design-based positions. Even a slight displacement from the design-based position of fuel rod may occur variation of flow area, and give influence upon the thermal hydraulic characteristics in the rod bundle. Therefore, if the critical power in the rod bundle is evaluated by an analytical approach, the analysis based on more correct input can be performed by using actual fuel rod position data. In this study, the rod positions in the rod bowing test section were measured using the high energy X-ray computer tomography (Xray-CT). Based on the measured rod positions data, the subchannel analysis by the NASCA code was performed, in order to investigate applicability of the NASCA code to BT estimation of the rod bowing test section, and influence of displacement from design-based rod position upon BT estimation by the NASCA code. The predicted critical powers are agreement with those obtained by the experiment. The analysis based on the design-based rod positions is also performed, and the result is

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  16. Investigation of an overheated PWR-type fuel rod simulator bundle cooled down by steam. Pt. 1: experimental and calculational results of the QUENCH-04 test. Pt. 2: application of the SVECHA/QUENCH code to the analysis of the QUENCH-01 and QUENCH-04 bundle tests

    International Nuclear Information System (INIS)

    Sepold, L.; Hofmann, P.; Homann, C.

    2002-04-01

    The QUENCH experiments are to investigate the hydrogen source term that results from the water injection into an uncovered core of a light-water reactor (LWR). The test bundle is made of 21 fuel rod simulators with a length of approximately 2.5 m. 20 fuel rod simulators are heated over a length of 1024 mm, the one unheated fuel rod simulator is located in the center of the test bundle. Heating is carried out electrically using 6-mm-diameter tungsten heating elements installed in the center of the rods and surrounded by annular ZrO 2 pellets. The rod cladding is identical to that used in LWRs: Zircaloy-4, 10.75 mm outside diameter, 0.725 mm wall thickness. The test bundle is instrumented with thermocouples attached to the cladding and the shroud at 17 different elevations with an axial distance between the thermocouples of 100 mm. During the entire test up to the cooldown phase, superheated steam together with the argon as carrier gas enters the test bundle at the bottom end and leaves the test section at the top together with the hydrogen that is produced in the zirconium-steam reaction. The hydrogen is analyzed by three different instruments: two mass spectrometers and a ''Caldos 7 G'' hydrogen measuring device (based on the principle of heat conductivity). Part I of this report describes the results of test QUENCH-04 performed in the QUENCH test facility at the Forschungszentrum Karlsruhe on June 30, 1999. The objective of the experiment QUENCH-04 was to investigate the reaction of the non-preoxidized rod cladding on cooldown by steam rather than quenching by water. Part II of the present report deals with the results of the SVECHA/QUENCH (S/Q) code application to the FZK QUENCH bundle tests. The adaptation of the S/Q code to such kind of calculations is described. The numerical procedure of the recalculation of the temperature test data, and the preparation for the S/Q code input is presented. In particular, the results of the QUENCH-01 and QUENCH-04 test

  17. Assessments of CHF correlations based on full-scale rod bundle experiments

    International Nuclear Information System (INIS)

    Sardh, K.; Becker, K.M.

    1986-02-01

    In the present study the Barnett, the Becker, the Biasi, the CISE-4, the XN-1, the EPRI and the Bezrukov burnout correlations have been compared with burnout measurements obtained with full scale 81, 64, 36 and 37-rod bundles. The total power as well as the local power hypothesis was employed for the comparisons. The results clearly indicated that the Biasi and the CISE-4 correlations do not predict the burnout conditions in full-scale rod bundles. Since, these correlations yield non-conservative results their use in computer programs as for instance RELAP, TRAC or NORA should be avoided. Considering that the effects of spacers were not included in the predictions, the Becker and the Bezrukov correlations were in excellent agreement with the experimental data. However, it should be pointed out that the Bezrukov correlation only covered the 70 and 90 bar data, while the Becker correlation agreed with the experimental data in the whole pressure range between 30 and 90 bar. The Barnett, the XN-1 and the EPRI correlations were also in satisfactory agreement with the experiments. We therefore conclude that for predictions of the burnout conditions in full-scale BWR rod bundles the Becker correlation should be employed. (author)

  18. A comparison of the accuracy of some correlations for burnout in annuli and rod bundles

    International Nuclear Information System (INIS)

    Barnett, P.G.

    1968-01-01

    The compilation of burnout data for annuli and rod bundles given in AEEW - R 463 is up-dated. A comparison is made of the accuracy with which five correlations for burnout in annuli predict the compiled annulus data; one correlation is found to be significantly better than the rest when the data is viewed as a whole. The accuracy with which this correlation and two of the other correlations predict the rod bundle burnout data is compared; the same correlation is again found to be the more likely to give accurate predictions. The implication of these observations is that it may be possible to make considerable savings in the cost of experimental and assessment work on rod bundles by studying their equivalent annuli. (author)

  19. Investigations with diagnostic fuel rod bundles on Rheinsberg NPP

    International Nuclear Information System (INIS)

    Krauze, F.; Rudolf, G.; Shajfler, V.; Tsimke, K.

    1982-01-01

    In 70MW pressurized water reactor of Rheinsberg NPP diagnostic fuel rod bundles have been installed: first of DK 1 type and then of DK 2 advanced type. Three rounds of measurement were run with DK 1 bundle and one with DK 2. The diagnostic bundles are equiped with various sensors for temperature, pressure, neutron flux and mechanical stress measurements as well as with special flow rate control system which allows to reach coolant boiling within the bundle. Qualitative and quantitative description of the sensors performance during reactor operation is given. The presented experimental results are connected with: 1) working capability of the measuring devices and their calibration; 2) throttling and boiling in two regimes: a) stationary and non-stationary flow rate throbgh DK during stationary reactor operation; b) various constant levels of flow rate through DK during non-stationary reactor operation regime [ru

  20. Analysis of fuel rod behaviour within a rod bundle of a pressurized water reactor under the conditions of a loss of coolant accident (LOCA) using probabilistic methodology

    International Nuclear Information System (INIS)

    Sengpiel, W.

    1980-12-01

    The assessment of fuel rod behaviour under PWR LOCA conditions aims at the evaluation of the peak cladding temperatures and the (final) maximum circumferential cladding strains. Moreover, the estimation of the amount of possible coolant channel blockages within a rod bundle is of special interest, as large coplanar clad strains of adjacent rods may result in strong local reductions of coolant channel areas. Coolant channel blockages of large radial extent may impair the long-term coolability of the corresponding rods. A model has been developed to describe these accident consequences using probabilistic methodology. This model is applied to study the behaviour of fuel rods under accident conditions following the double-ended pipe rupture between collant pump and pressure vessel in the primary system of a 1300 MW(el)-PWR. Specifically a rod bundle is considered consisting of 236 fuel rods, that is subjected to severe thermal and mechanical loading. The results obtained indicate that plastic clad deformations with circumferential clad strains of more than 30% cannot be excluded for hot rods of the reference bundle. However, coplanar coolant channel blockages of significant extent seem to be probable within that bundle only under certain boundary conditions which are assumed to be pessimistic. (orig./RW) [de

  1. Fuel rod bundles proposed for advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Prodea, Iosif; Catana, Alexandru

    2010-01-01

    The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

  2. Development of multidimensional two-phase flow measurement sensor in rod bundle

    International Nuclear Information System (INIS)

    Arai, Takahiro; Furuya, Masahiro; Shirakawa, Kenetsu; Kanai, Taizo

    2011-01-01

    In order to acquire multidimensional two-phase flow in 10x10 bundle, SubChannel Void Sensor (SCVC) consisting of 11-wire by 11-wire and 10-rod by 10-rod electrodes is developed. A conductance value in a proximity region of one wire and another gives void fraction in the center of subchannel region. A phasic velocity can be estimated by using two layers of wire meshes, like as so-called wire mesh sensor. 121 points (=11x11) of void fraction as well as those of phasic velocity are acquired. It is peculiarity of the devised sensor that void fraction near rod surface can be estimated by a conductance value in a proximity region of one wire and one rod. 400 additional points of void fraction in 10x10 bundle can be, therefore, acquired. The time resolution of measurement is up to 1250 frames (cross sections) per second. We capability in a 10x10 bundle with o.d. 10 mm and 3110 mm long is demonstrated. The devised sensor is installed in 8 height levels to acquire the two-phase flow dynamics along axial direction. A pair of sensor layers is mounted in each level and is placed by 30 mm apart with each other to estimate a phasic velocity distribution on the basis of cross-correlation function of the two layers. Air bubbles are injected through sintered metal nozzles from the bottom end of 10x10 rods. Air flow rate distribution can vary with a controlled valves connected to each nozzle. The devised sensor exhibited the quasi three-dimensional flow structures, i.e. void fraction, phasic velocity and bubble chord length distributions. These quasi three-dimensional structures explorer complexity of two-phase flow dynamics such as coalescence and breakup of bubbles in the transient phasic velocity distributions. (author)

  3. Studies of decay heat removal by natural convection using the SONACO sodium-cooled 37-pin bundle

    International Nuclear Information System (INIS)

    Wydler, P.; Dury, T.V.; Hudina, M.; Weissenfluh, T. von; Sigg, B.; Dutton, P.

    1986-01-01

    Natural convection measurements in an electrically heated sodium-cooled rod bundle are being performed with the aim of contributing to a better understanding of natural convection effects in subassemblies with stagnant sodium and providing data for code validation. Measurements include temperature distributions in the bundle for different cooling configurations which simulate heat transfer to the intersubassembly gap and neighbouring subassemblies and possible thermosyphonic interaction between a subassembly and the reactor plenum above. Conditions for which stable natural convection patterns exist are identified, and results are compared with predictions of different computer codes of the porous-medium type. (author)

  4. Analytical prediction of friction factors and Nusselt numbers of turbulent forced convection in rod bundles with smooth and rough surfaces

    International Nuclear Information System (INIS)

    Su Jian; Silva Freire, Atila P.

    2002-01-01

    A simple analytical method was developed for the prediction of the friction factor, f, of fully developed turbulent flow and the Nusselt number, Nu, of fully developed turbulent forced convection in rod bundles arranged in square or hexagonal arrays. The friction factor equation for smooth rod bundles was presented in a form similar to the friction factor equation for turbulent flow in a circular pipe. An explicit equation for the Nusselt number of turbulent forced convection in rod bundles with smooth surface was developed. In addition, we extended the analysis to rod bundles with rough surface and provided a method for the prediction of the friction factor and the Nusselt number. The method was based on the law of the wall for velocity and the law of the wall for the temperature, which were integrated over the entire flow area to yield algebraic equations for the prediction of f and Nu. The present method is applicable to infinite rod bundles in square and hexagonal arrays with low pitch to rod diameter ratio, P/D<1.2

  5. Assessment of a non-uniform heat flux correction model to predicting CHF in PWR rod bundles

    International Nuclear Information System (INIS)

    Dae-Hyun, Hwang; Sung-Quun, Zee

    2001-01-01

    author for the prediction of CHF in a boiling channel with nonuniform axial heat flux distributions. In this study, we assess the applicability of the proposed model for PWR rod bundles. (authors)

  6. Numerical simulation of bubble motion about a grid spacer in a rod bundle

    International Nuclear Information System (INIS)

    Zhang, Zheng; Hosokawa, Shigeo; Hayashi, Kosuke; Tomiyama, Akio

    2009-01-01

    Numerical simulations based on a three-dimensional two-way bubble tracking method are carried out to predict bubble motions in a square duct with an obstacle and in a two-by-three rod bundle with a grid spacer. Comparisons between measured and predicted bubble motions demonstrate that the two-way bubble tracking method gives good predictions for trajectories of small bubbles in the upstream side of the grid spacer in the rod bundle geometry. The predicted bubble trajectories clearly show that bubbles are apt to migrate toward the rod surface in the vicinity of the bottom of the grid spacer. Analysis of forces acting on the bubbles confirms that pressure gradient force induced by the presence of the spacer is the main cause of the bubble lateral migration toward the rod surface. Motions of steam bubbles at a nominal operating condition of a pressurized water reactor (PWR) are also predicted by using the bubble tracking method, which indicates that steam bubbles also migrate toward the rod surface at the upstream side of the spacer due to the spacer-induced pressure gradient force. (author)

  7. Measurement of liquid film flow on nuclear rod bundle in micro-scale by using very high speed camera system

    Science.gov (United States)

    Pham, Son; Kawara, Zensaku; Yokomine, Takehiko; Kunugi, Tomoaki

    2012-11-01

    Playing important roles in the mass and heat transfer as well as the safety of boiling water reactor, the liquid film flow on nuclear fuel rods has been studied by different measurement techniques such as ultrasonic transmission, conductivity probe, etc. Obtained experimental data of this annular two-phase flow, however, are still not enough to construct the physical model for critical heat flux analysis especially at the micro-scale. Remain problems are mainly caused by complicated geometry of fuel rod bundles, high velocity and very unstable interface behavior of liquid and gas flow. To get over these difficulties, a new approach using a very high speed digital camera system has been introduced in this work. The test section simulating a 3×3 rectangular rod bundle was made of acrylic to allow a full optical observation of the camera. Image data were taken through Cassegrain optical system to maintain the spatiotemporal resolution up to 7 μm and 20 μs. The results included not only the real-time visual information of flow patterns, but also the quantitative data such as liquid film thickness, the droplets' size and speed distributions, and the tilt angle of wavy surfaces. These databases could contribute to the development of a new model for the annular two-phase flow. Partly supported by the Global Center of Excellence (G-COE) program (J-051) of MEXT, Japan.

  8. Performance assessment of the RANS turbulence models in nuclear fuel rod bundles

    International Nuclear Information System (INIS)

    In, Wang Kee; Chun, Tae Hyun; Oh, Dong Seok; Shin, Chang Hwan

    2005-02-01

    The three experiments for turbulent flow in a rod bundle geometry were simulated in this CFD analysis using various RANS models. The CFD predictions were compared with the experimental and DNS results. The RANS models used here are the nonlinear quadratic/cubic κ-ε models and the second-order closure models (SSG, LRR, RSM-ω). The anisotropic models predicted the secondary flow and showed a significantly improved agreement with the measurements from the standard κ-ε model. In particular, the SSG model resulted in the best performance showing the closest agreement with the experimental results. However, the RANS models could not predict the very high anisotropy observed in a rod bundle with a small pitch-to-diameter ratio

  9. SEFLEX - fuel rod simulator effects in flooding experiments. Pt. 2

    International Nuclear Information System (INIS)

    Ihle, P.; Rust, K.

    1986-03-01

    This report presents typical data and a limited heat transfer analysis from unblocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5 x 5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5 x 5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

  10. SEFLEX fuel rod simulator effects in flooding experiments. Pt. 3

    International Nuclear Information System (INIS)

    Ihle, P.; Rust, K.

    1986-03-01

    This report presents typical data and a limited heat transfer analysis from blocked bundle reflood tests of an experimental thermal-hydraulic program. Full-length bundles of 5x5 fuel rod simulators having a gas-filled gap between the Zy cladding and the alumina pellets were tested in the test rig designed for the earlier Flooding Experiments with Blocked Arrays (FEBA-program). The 5x5 FEBA rod bundle tests were performed with gapless heater rods. These rods have a close thermal contact between the stainless steel cladding and the electric insulation material. A comparison of the SEFLEX data with the reference data of FEBA obtained under identical initial and reflood conditions shows the influence of different fuel rod simulators on the thermal-hydraulic behavior during forced feed bottom reflooding of unblocked and blocked arrays. Compared to bundles of gapless rods, bundles of rods with Zy claddings and a gas filled gap between claddings and pellets, which more closely represent the features that exist in an actual fuel rod geometry, produced higher quench front velocities, enhanced removal of stored heat in the rods, reduced peak cladding temperatures, increased grid spacer effects and absolutely unproblematic coolability of 90 percent blockages with bypass. The data offer the opportunity for further validation of computer codes to make realistic predictions of safety margins during a LOCA in a PWR. (orig./HP) [de

  11. Measurements of two-phase flow patterns in a 4 x 4 rod bundle

    International Nuclear Information System (INIS)

    Akio tomiyama; Akira Sou; Shigeo Hosokawa; Masato Mitsuhashi; Kohei Noda; Yasushi Tsubo; Kaichiro Mishima; Yoshiro Kudo

    2005-01-01

    Air-water two-phase flow patterns in a 4 x 4 square lattice rod bundle consisting of an acrylic channel box of 68 mm in width and transparent rods of 12 mm in diameter were measured by utilizing FEP (fluorinated ethylene propylene) tubes for the rods. The FEP possesses the same refractive index with water, and therefore, whole flow patterns in the bundle and local flow patterns in subchannels were visualized with little optical distortion. In addition to the visualization, transmission rates of laser beam from one rod to its opponent rod and two-point correlation coefficients of phase indicator functions were measured to examine the feasibility of objective identification of flow patterns in subchannels. The ranges of liquid and gas volume fluxes, JL and JG, were 0.1 < JL < 2.0 m/s and 0.04 < JG < 8.85 m/s, respectively. As a result, the following conclusions were obtained: (1) slug flow pattern does not appear in the rod bundle and bubbly flow would directly transit to churn flow, (2) the measured boundary between bubbly and churn flows is close to the boundary between bubbly and slug flows given by Mishima and Ishii's flow pattern transition model, (3) critical void fraction causing bubbly to churn flow transition depends on a subchannel, i.e., about 0.3 for inner subchannels, about 0.2 for side subchannels and about 0.1 for corner subchannels, and (4) the two-point correlation coefficient of phase indicator functions for two inner subchannels shows a steep increase at the bubbly to churn flow transition, which, in turn, means that the two-point correlation is an appropriate indicator for detecting this transition. (authors)

  12. Multicell slug flow heat transfer analysis of finite LMFBR bundles

    International Nuclear Information System (INIS)

    Yeung, M.K.; Wolf, L.

    1978-12-01

    An analytical two-dimensional, multi-region, multi-cell technique has been developed for the thermal analysis of LMFBR rod bundles. Local temperature fields of various unit cells were obtained for 7, 19, and 37-rod bundles of different geometries and power distributions. The validity of the technique has been verified by its excellent agreement with the THTB calculational result. By comparing the calculated fully-developed circumferential clad temperature distribution with those of the experimental measurements, an axial correction factor has been derived to account for the entrance effect for practical considerations. Moreover, the knowledge of the local temperature field of the rod bundle leads to the determination of the effective mixing lengths L/sub ij/ for adjacent subchannels of various geometries. It was shown that the implementation of the accurately determined L/sub ij/ into COBRA-IIIC calculations has fairly significant effects on intersubchannel mixing. In addition, a scheme has been proposed to couple the 2-D distributed and lumped parameter calculation by COBRA-IIIC such that the entrance effect can be implanted into the distributed parameter analysis. The technique has demonstrated its applicability for a 7-rod bundle and the results of calculation were compared to those of three-dimensional analyses and experimental measurements

  13. Friction Factors in Rough Rod Bundles Estimated from Experiments in Partially Rough Annuli - Effects of Dissimilarities in the Shear Stress and Turbulence Distributions

    Energy Technology Data Exchange (ETDEWEB)

    Kjellstroem, B

    1968-12-15

    Experiments with rough surface friction and heat transfer are often made in an annulus with rough inner surface and smooth outer surface. Utilization of data from such experiments for calculation of rough rod bundle fuel elements requires a transformation of the data. For this purpose the method of WB Hall is frequently used. The errors introduced by two of the assumptions on which this method is based, namely the assumptions of zero shear at the radius of maximum velocity and the assumption of no turbulence exchange between the subchannels, are discussed, and the magnitude of the errors is estimated on basis of experiments in a partially rough annulus. It is found that the necessary corrections does not amount to more than about + 10 % for the friction factor and + 15 % for the Reynolds number and the equivalent diameter. The correction for the turbulence exchange alone is of the order of 2-3 %. A comparison of friction factors measured in a rough 48-rod bundle and predicted from measurements in a partially rough annulus was also made. The prediction was 5 % high instead of about 10 % low which could have been expected from the considerations earlier in the report. Explanations for this can be found in the effect of the channel shape or inaccuracies in the rod bundle experiment. Annulus experiments which will allow comparisons with other rod bundle experiments will be run to clarify this.

  14. Experimental study of mixing in a square array rod bundle with grid spacer

    International Nuclear Information System (INIS)

    Zong Guifang; Cai Zuti; Zhang Demei

    1989-01-01

    This paper describes the experimental study of mixing in a full scale 15x15 square array rod bundle fuel assembly with 10 mm diameter and 13.3 mm pitch. The experiment was carried out in an open water loop, K 2 CrO 4 was used as tracer. Each subchannel was sampled at the open bundle outlet. Titration, spectrophotometry and fibreoptic methods were used to measure the concentration. The Reynolds numbers ranged from 2.12x10 4 to 4.37x10 4 . For the turbulent mixing of the bare rod bundle, the results of this study agreed with the formulas recommended by other authors. Both flow visualisation studies and the quantitative analysis indicated that flow scattering caused by the grid has a little effect on the mixing. The cause has been examined in this paper. (orig.)

  15. Measurement and model development of the droplet diameter in rod bundles with spacer grids in the reactor core

    Energy Technology Data Exchange (ETDEWEB)

    No, Hee Cheon; Lee, Eo Hwak; Yoo, Seung Hun; Jin, Hyung Gon; Kim, In Hun [KAIST, Daejeon (Korea, Republic of)

    2010-05-15

    To understand and to predict the heat transfer between superheated steam and droplets properly during reflood phase of LBLOCA of APR1400, it is very important to measure broken droplet sizes by spacer grids. A study, therefore, has been performed to investigate droplet size in rod bundles with spacer grids and to develop a spacer grid droplet size model for safety analysis codes. Experiments were conducted with liquid droplets (SMD of 300{approx}700 {mu}m) impacting on various spacer grids at air superficial velocity of 10 and 20 m/s based on FLECHT SEASET. The test channel and the grids were heated to 150 .deg. C to prevent the formation of liquid film during tests. The spacer grids were designed refer to the Korean fuel rod bundles (Korean Standard Fuel, Plus 7) of APR1400 with various blockage area ratio and grid geometries (strap thickness, mixing vane) and about 15,000 droplets were measured at upstream and downstream of the grids in 16 tests. As a result, the measurement of broken droplet size by spacer grids with photography method is presented and the droplet size model related to spacer grids as a function of blockage area ratio is suggested in this report

  16. Semi-empirical model for the calculation of flow friction factors in wire-wrapped rod bundles

    International Nuclear Information System (INIS)

    Carajilescov, P.; Fernandez y Fernandez, E.

    1981-08-01

    LMFBR fuel elements consist of wire-wrapped rod bundles, with triangular array, with the fluid flowing parallel to the rods. A semi-empirical model is developed in order to obtain the average bundle friction factor, as well as the friction factor for each subchannel. The model also calculates the flow distribution factors. The results are compared to experimental data for geometrical parameters in the range: P(div)D = 1.063 - 1.417, H(div)D = 4 - 50, and are considered satisfactory. (Author) [pt

  17. Transient non-boiling heat transfer in a fuel rod bundle during accidental power excursions

    International Nuclear Information System (INIS)

    Bonaekdarzadeh, S.; Johannsen, K.; Ramm, H.

    1977-01-01

    The physical problem studied is the transient non-boiling heat transfer of a cylindrical fuel rod consisting of fuel, gap, and cladding to a steady, fully developed turbulent flow. The fuel pin is assumed to be located in the interior region of a subassembly with regular triangular or square arrangements. The turbulent velocity field as well as turbulent transport properties are specified as functions of the coordinates normal to the axial flow direction. The heat generation within the fuel may be specified as an arbitrary function of the three spatial coordinates and time. A digital computer program has been developed. On the basis of finite-difference techniques, to solve the governing partial differential equations with their associated subsidiary conditions. Results have been obtained for a series of exponential power transients of interest to safety of liquid-metal and water cooled nuclear reactors. The general physical features of transient convective heat transfer as explored by previous investigators have qualitatively been substantiated by the present analysis. Emphasis has been devoted to investigate the differences of heat-transfer (coefficient) results from multi-region analysis including a realistic fuel rod model and single-region analysis for the coolant region only. A comparison with the engineering relationships for turbulent liquid-metal cooling by Stein, which are an extension of the heat transfer coefficient concept to account for transient heat fluxes, clearly demonstrates that, at the parameters studied, Stein's approach tends to largely overestimate the convective heat transfer at early times

  18. Determination of mass flow rate and quality distributions between the subchannels of a heated bundle. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bayoumi, M.; Charlot, R.; Ricque, R.

    1976-05-01

    For analyzing, correlating and extrapolating experimental burn-out results obtained with LWR rod bundles, it is necessary to know the distributions of mass flow rate and quality between the subchannels. A description is presented of an experimental study in progress at the CEN-Grenoble for determining and adjusting the laws of mixing in the FLICA Code which is used to predict these distributions. The experiments are performed on the FRENESIE loop with Freon 12. The test section, in vertical position, consists of a four rod bundle in a channel with square section. The heat flux is axially uniform. The flow of each subchannel can be sampled in ''isokinetic conditions,'' at the end of the heating length. Thermodynamic quality and mass flow rate of the samplings are measured in steady state conditions by using respectively a calorimeter and a turbine flow meter. The test facility is described and experimental data are presented and discussed.

  19. Development of subchannel void measurement sensor and multidimensional two-phase flow dynamics in rod bundle

    International Nuclear Information System (INIS)

    Arai, T.; Furuya, M.; Kanai, T.; Shirakawa, K.

    2011-01-01

    An accurate subchannel database is crucial for modeling the multidimensional two-phase flow in a rod bundle and for validating subchannel analysis codes. Based on available reference, it can be said that a point-measurement sensor for acquiring void fractions and bubble velocity distributions do not infer interactions of the subchannel flow dynamics, such as a cross flow and flow distribution, etc. In order to acquire multidimensional two-phase flow in a 10×10 rod bundle with an o.d. of 10 mm and 3110 mm length, a new sensor consisting of 11-wire by 11-wire and 10-rod by 10-rod electrodes was developed. Electric potential in the proximity region between two wires creates a void fraction in the center subchannel region, like a so-called wire mesh sensor. A unique aspect of the devised sensor is that the void fraction near the rod surface can be estimated from the electric potential in the proximity region between one wire and one rod. The additional 400 points of void fraction and phasic velocity in 10×10 bundle can therefore be acquired. The devised sensor exhibits the quasi three-dimensional flow structures, i.e. void fraction, phasic velocity and bubble chord length distributions. These quasi three-dimensional structures exhibit the complexity of two-phase flow dynamics, such as coalescence and the breakup of bubbles in transient phasic velocity distributions. (author)

  20. Experimental investigations on the fluid flow through an asymmetric rod bundle (W/D = 1.026)

    International Nuclear Information System (INIS)

    Rehme, K.

    1982-05-01

    Measurements of the distributions of the mean velocity, the wall shear stresses and the turbulence were performed in a wall subchannel of a rod bundle of four parallel rods arranged asymmetrically in a rectangular channel (P/D = 1.07, W/D = 1.026). The Reynolds number of this investigation was Re = 5.46 x 10 4 . The experimental results show that the momentum transport is highly anisotropic especially in the gaps of the rod bundle. Influences of secondary flow cannot be detected in the distribution of the time-mean velocity, however, such influences are found in the distributions of the turbulence intensities and the kinetic energy of turbulence. The comparison between experimental wall shear stress distributions and those calculated with the VELASCO-code shows discrepancies especially in the gap between the rod and channel walls. (orig.) [de

  1. Single-phase CFD applicability for estimating fluid hot-spot locations in a 5 x 5 fuel rod bundle

    International Nuclear Information System (INIS)

    Ikeda, Kazuo; Makino, Yasushi; Hoshi, Masaya

    2006-01-01

    High-thermal performance PWR spacer grids require both of low pressure loss and high critical heat flux (CHF) properties. Therefore, a numerical study using computational fluid dynamics (CFD) was carried out to estimate pressure loss in strap and mixing vane structures. Moreover, a CFD simulation under single-phase flow condition was conducted for one specific condition in a water departure from nucleate boiling (DNB) test to examine the applicability of the CFD model for predicting the CHF rod position. Energy flux around the rod surface in a water DNB test is the sum of the intrinsic energy flux from a rod and the extrinsic energy flux from other rods, and increments of the enthalpy and decrements of flow velocity near the rod surface are assumed to affect CHF performance. CFD makes it possible to model the complicated flow field consisting of a spacer grid and a rod bundle and evaluate the local velocity and enthalpy distribution around the rod surface, which are assumed to determine the initial conditions for the two-phase structure. The results of this study indicate that single-phase CFD can play a significant role in designing PWR spacer grids for improved CHF performance

  2. Investigation of combined free and forced convection in a 2 x 6 rod bundle during controlled flow transients

    International Nuclear Information System (INIS)

    Bates, J.M.; Khan, E.U.

    1980-10-01

    An experimental study was performed to obtain local fluid velocity and temperature measurements in the mixed (combined free and forced) convection regime for specific flow coastdown transients. A brief investigation of steady-state flows for the purely free-convection regime was also completed. The study was performed using an electrically heated 2 x 6 rod bundle contained in a flow housing. In addition a transient data base was obtained for evaluating the COBRA-WC thermal-hydraulic computer program

  3. Single-phase and two-phase gas-liquid turbulent mixing between subchannels in a simulated rod bundle

    International Nuclear Information System (INIS)

    Sadatomi, Michio; Kawahara, Akimaro; Sato, Yoshifusa; Tomino, Takayoshi.

    1996-01-01

    This study is concerned with turbulent mixing which is one of the three mechanisms of cross flows between subchannels in a nuclear fuel rod bundle. The channel used in this experiments was a vertical simulated rod bundle having two subchannels connected through 1 to 3 gaps between two rods and/or rod and channel wall. The number of the gaps was changed to investigate the effect of the number on the turbulent mixing. Turbulent mixing rates of air and water and fluctuations of pressure difference between the subchannels were measured for single-phase and two-phase gas-liquid flows under hydrodynamic equilibrium flow conditions. It has been confirmed that the turbulent mixing rate is affected strongly by the fluctuations especially for liquid phase in two-phase slug or churn flow. (author)

  4. Effects of duct configuration on flow and temperature structure in sodium-cooled 19-rod simulated LMFBR fuel bundles with helical wire-wrap spacers

    International Nuclear Information System (INIS)

    Wantland, J.L.; Fontana, M.H.; Gnadt, P.A.; Hanus, N.; MacPherson, R.E.; Smith, C.M.

    1976-01-01

    Thermal-hydrodynamic testing of sodium-cooled 19-rod simulated LMFBR fuel bundles is being conducted at the O ak Ridge National Laboratory in the Fuel Failure Mockup (FFM), an engineering-scale high-temperature sodium facility which provides prototypic flows, temperatures and power densities. Electrically heated bundles have been tested with two scalloped and two hexagonal duct configurations. Peripheral helical flows, attributed to the spacers, have been observed with strengths dependent upon the evenness and relative sizes of the peripheral flow areas. Diametral sodium temperature profiles are more uniform with smaller peripheral flow areas

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

    International Nuclear Information System (INIS)

    Hisashi, Ninokata

    2006-01-01

    In order to practice a design-by-analysis of thermohydraulics design of BWR fuel rod bundles, the subchannel analysis would play a major role. There, the immediate concern is improvement in its predictive capability of CHF due in particular to the film dryout (boiling transition phenomena: BT) on the fuel rod surface. Constitutive equations in the subchannel analysis formulation are responsible for the quality of calculated results. The constitutive equations are a result of integration of the local and instantaneous description of two-phase flows over the subchannel control volume. In general, they are expressed in terms of subchannel-control-volume- as well as area-averaged two-phase flow state variables. In principle the information on local and instantaneous physical phenomena taking place inside subchannels must be counted for in the algebraic form of the equations on the basis of a more mechanistic modeling approach. They should include also influences of the multi-dimensional subchannel geometry and fluid material properties. Thermohydraulics phenomena of interests in this deed are: 1) vapor-liquid re-distribution by inter-subchannel exchanges due to the diversion cross flow, turbulent mixing and void drift, 2) liquid film behaviors, 3) transition of two-phase flow regimes, 4) droplet entrainment and deposition and 5) spacer-droplet interactions. These are considered to be five key factors in understanding the BT in BWR fuel rod bundles. In Japan, a university-industry consortium has been formed under the sponsorship of the Ministry of Economics, Trade and Industry. This paper describes an outline of the on-going project and, first, an outline of the current efforts is presented in developing a new two-fluid three field subchannel code NASCA being aimed at predicting onset of BT, and post BT phenomena in advanced BWR fuel rod bundles including those of the tight lattice configuration for a higher conversion. Then the current methodology adopted to improve

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

    International Nuclear Information System (INIS)

    Hisashi, Ninokata

    2004-01-01

    Full text of publication follows:In order to practice a design-by-analysis of thermohydraulics design of BWR fuel rod bundles, the subchannel analysis would play a major role. There, the immediate concern is improvement in its predictive capability of CHF due in particular to the film dryout (boiling transition phenomena: BT) on the fuel rod surface. Constitutive equations in the subchannel analysis formulation are responsible for the quality of calculated results. The constitutive equations are a result of integration of the local and instantaneous description of two-phase flows over the subchannel control volume. In general, they are expressed in terms of subchannel-control-volume- as well as area-averaged two-phase flow state variables. In principle the information on local and instantaneous physical phenomena taking place inside subchannels must be counted for in the algebraic form of the equations on the basis of a more mechanistic modeling approach. They should include also influences of the multi-dimensional subchannel geometry and fluid material properties. Thermohydraulics phenomena of interests in this deed are: 1) vapor-liquid re-distribution by inter-subchannel exchanges due to the diversion cross flow, turbulent mixing and void drift, 2) liquid film behaviors, 3) transition of two-phase flow regimes, 4) droplet entrainment and deposition and 5) spacer-droplet interactions. These are considered to be five key factors in understanding the BT in BWR fuel rod bundles. In Japan, a university-industry consortium has been formed under the sponsorship of the Ministry of Economics, Trade and Industry. This paper describes an outline of the on-going project and, first, an outline of the current efforts is presented in developing a new two-fluid three field subchannel code NASCA being aimed at predicting onset of BT, and post BT phenomena in advanced BWR fuel rod bundles including those of the tight lattice configuration for a higher conversion. Then the current

  7. Algebraic stress model for axial flow in a bare rod-bundle

    International Nuclear Information System (INIS)

    de Lemos, M.J.S.

    1987-01-01

    The problem of predicting transport properties for momentum and heat across the boundaries of interconnected channels has been the subject of many investigations. In the particular case of axial flow through rod-bundles, transport coefficients for channel faces aligned with rod centers are known to be considerably higher than those calculated by simple isotropic theories. And yet, it was been found that secondary flows play only a minor role in this overall transport, being turbulence highly enhanced across that hypothetical surface. In order to numerically predict the correct amount of the quantity being transported, the approach taken by many investigators was then to artificially increase the diffusion coefficient obtained via a simple isopropic theory (usually the standard k-ε model) and numerically match the correct experimentally observed mixing rates. The present paper reports an attempt to describe the turbulent stresses by means of an Algebraic Stress Model for turbulence. Relative turbulent kinetic energy distribution in all three directions are presented and compared with experiments in a square lattice. The strong directional dependence of transport terms are then obtained via a model for the Reynolds stresses. The results identify a need for a better representation of the mean-flow field part of the pressure-strain correlation term

  8. Model for transversal turbulent mixing in axial flow in rod bundles

    International Nuclear Information System (INIS)

    Carajilescov, P.

    1990-01-01

    The present work consists in the development of a model for the transversal eddy diffusivity to account for the effect of turbulent thermal mixing in axial flows in rod bundles. The results were compared to existing correlations that are currently being used in reactor thermalhydraulic analysis and considered satisfactory. (author)

  9. Large bundle BWR test CORA-18: Test results

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Noack, V.; Sepold, L.; Schanz, G.; Schumacher, G.

    1998-04-01

    The CORA out-of-pile experiments are part of the international Severe Fuel Damage (SFD) Program. They were performed to provide information on the damage progression of Light Water Reactor (LWR) fuel elements in Loss-of-coolant Accidents in the temperature range 1200 C to 2400 C. CORA-18 was the large BWR bundle test corresponding to the PWR test CORA-7. It should investigate if there exists an influence of the BWR bundle size on the fuel damage behaviour. Therefore, the standard-type BWR CORA bundle with 18 fuel rod simulators was replaced by a large bundle with two additional surrounding rows of 30 rods (48 rods total). Power input and steam flow were increased proportionally to the number of fuel rod simulators to give the same initial heat-up rate of about 1 K/s as in the smaller bundles. Emphasis was put on the initial phase of the damage progression. More information on the chemical composition of initial and intermediate interaction products and their relocation behaviour should be obtained. Therefore, power and steam input were terminated after the onset of the temperature escalation. (orig.) [de

  10. CFD simulation of crossflow mixing in a rod bundle with mixing blades

    International Nuclear Information System (INIS)

    In, W. K.

    1999-01-01

    A CFD model was developed in this study to simulate the crossflow mixing in a 4x4 square array rod bundle caused by ripped-open blades. The central subchannel and adjacent subchannels of one grid span were modeled using flow symmetry. The lateral velocity pattern within the central subchannel, lateral velocity and the turbulence intensity in the rod gap region were predicted by the CFD method, and the predictions were compared with the measurements. The CFD simulation shows a vortex flow around the fuel rod caused by a pair of blades, which is consistent with the experimental results. The CFD predictions of the lateral velocity on the mixing sections show a near symmetric profile, but the measurements present an asymmetric velocity profile leading to an inversion of lateral velocity. The predicted mixing rate between the central subchannel and the adjacent subchannels reasonably agrees with the measured one. The CFD prediction shows a parabolic distribution of the RMS velocity but the measured one shows a rather flat distribution near the blade that develops to a parabolic distribution far downstream (L=29De). The predicted average RMS velocity on a mixing section is also slightly lower than the measured one. This study confirmed that the CFD simulation can present the effect of the ripped-open blades on the crossflow mixing in a rod bundle well

  11. Local heat transfer where heated rods touch in axially flowing water

    International Nuclear Information System (INIS)

    Kast, S.J.

    1983-05-01

    An anlaytic model is developed to predict the azimuthal width of a stablesteam blanket region near the line of contact between two heated rods cooled by axially flowing water at high pressure. The model is intended to aid analysis of reduced surface heat transfer capability for the abnormal configuration of nuclear fuel rods bowed into contact in the core of a pressurized water nuclear reactor. The analytic model predicts the azimuthal width of the steam blanket zone having reduced surface heat transfer as a function of rod average heat flux, subchannel coolant conditions and rod dimensions. The analytic model is developed from a heat balance between the heat generated in the wall of a heated empty tube and the heat transported away by transverse mixing and axial convection in the coolant subchannel. The model is developed for seveal geometries including heated rods in line contact, a heated rod touching a short insulating plane and a heated rod touching the inside of a metal guide tube

  12. REBEKA bundle experiments

    International Nuclear Information System (INIS)

    Wiehr, K.

    1988-05-01

    This report is a summary of experimental investigations describing the fuel rod behavior in the refilling and reflooding phase of a loss-of-coolant accident of a PWR. The experiments were performed with 5x5 and 7x7 rod bundles, using indirectly electrically heated fuel rod simulators of full length with original PWR-KWU-geometry, original grid spacers and Zircaloy-4-claddings (Type Biblis B). The fuel rod simulators showed a cosine shaped axial power profile in 7 steps and continuous, respectively. The results describe the influence of the different parameters such as bundle size on the maximum coolant channel blockage, that of the cooling on the size of the circumferential strain of the cladding (azimuthal temperature distribution) a cold control rod guide thimble and the flow direction (axial temperature distribution) on the resulting coolant channel blockage. The rewetting behavior of different fuel rod simulators including ballooned and burst Zircaloy claddings is discussed as well as the influence of thermocouples on the cladding temperature history and the rewetting behavior. All results prove the coolability of a PWR in the case of a LOCA. Therefore, it can be concluded that the ECC-criteria established by licensing authorities can be fulfilled. (orig./HP) [de

  13. Heat Transfer Analysis in Wire Bundles for Aerospace Vehicles

    Science.gov (United States)

    Rickman, S. L.; Iamello, C. J.

    2016-01-01

    Design of wiring for aerospace vehicles relies on an understanding of "ampacity" which refers to the current carrying capacity of wires, either, individually or in wire bundles. Designers rely on standards to derate allowable current flow to prevent exceedance of wire temperature limits due to resistive heat dissipation within the wires or wire bundles. These standards often add considerable margin and are based on empirical data. Commercial providers are taking an aggressive approach to wire sizing which challenges the conventional wisdom of the established standards. Thermal modelling of wire bundles may offer significant mass reduction in a system if the technique can be generalized to produce reliable temperature predictions for arbitrary bundle configurations. Thermal analysis has been applied to the problem of wire bundles wherein any or all of the wires within the bundle may carry current. Wire bundles present analytical challenges because the heat transfer path from conductors internal to the bundle is tortuous, relying on internal radiation and thermal interface conductance to move the heat from within the bundle to the external jacket where it can be carried away by convective and radiative heat transfer. The problem is further complicated by the dependence of wire electrical resistivity on temperature. Reduced heat transfer out of the bundle leads to higher conductor temperatures and, hence, increased resistive heat dissipation. Development of a generalized wire bundle thermal model is presented and compared with test data. The steady state heat balance for a single wire is derived and extended to the bundle configuration. The generalized model includes the effects of temperature varying resistance, internal radiation and thermal interface conductance, external radiation and temperature varying convective relief from the free surface. The sensitivity of the response to uncertainties in key model parameters is explored using Monte Carlo analysis.

  14. Experimental investigation of the turbulent flow through a wall subchannel of a rod bundle

    International Nuclear Information System (INIS)

    Rehme, K.

    1977-04-01

    An experimental investigation was performed to establish reliable information on the transport properties of turbulent flow through subchannels of rod bundles. Detailed data were measured of the distributions of the time-mean velocity, the turbulence intensities in all directions and, thus, the kinetic energy of turbulence, of the shear stresses in the directions normal and parallel to the walls, and of the wall shear stresses for a wall subchannel of a rod bundle of four rods in parallel. The pitch-to-diameter ratio of the rods equal to the wall-to-diameter ratio was 1.07, the Reynolds number of this investigation was Re = 8.7 x 10 4 . On the basis of the data measured the eddy viscosities in the directions normal and parallel to the walls were calculated. Thus, detailed data of the eddy viscosities in direction parallel to the walls in rod bundels were obtained for the first time. The experimental results were compared with predictions by the VELASCO-code. There are considerable differences between calculated and measured data of the time-mean velocity and the wall shear stresses. Attempts to adjust the VELASCO-code against the measurements were not successful. The reasons of the discrepancies are discussed. (orig.) [de

  15. Experimental investigation of the coolability of blocked hexagonal bundles

    Energy Technology Data Exchange (ETDEWEB)

    Hózer, Zoltán, E-mail: zoltan.hozer@energia.mta.hu; Nagy, Imre; Kunstár, Mihály; Szabó, Péter; Vér, Nóra; Farkas, Róbert; Trosztel, István; Vimi, András

    2017-06-15

    Highlights: • Experiments were performed with electrically heated hexagonal fuel bundles. • Coolability of ballooned VVER-440 type bundle was confirmed up to high blockage rate. • Pellet relocation effect causes delay in the cool-down of the bundle. • The bypass line does not prevent the reflood of ballooned fuel rods. - Abstract: The CODEX-COOL experimental series was carried out in order to evaluate the effect of ballooning and pellet relocation in hexagonal bundles on the coolability of fuel rods after a LOCA event. The effects of blockage geometry, coolant flowrate, initial temperature and axial profile were investigated. The experimental results confirmed that a VVER bundle up to 80% blockage rate remains coolable after a LOCA event under design basis conditions. The ballooned section creates some obstacles for the cooling water during reflood of the bundle, but this effect causes only a short delay in the cooling down of the hot fuel rods. The accumulation of fuel pellet debris in the ballooned volume results in a local power peak, which leads to further slowing down of quench front.

  16. Experimental study of the phenomena of turbulent flow in the narrow gaps between subchannels of rod bundles

    International Nuclear Information System (INIS)

    Moeller, S.V.

    1989-01-01

    It was observed that the turbulent intensities in the narrow gaps between the subchannels of rod bundles are strongly anisotropic and higher than in pipes. In rod bundles, both the axial and azimuthal components of the fluctuating velocity have a quasi-periodic behaviour. The intensities increase with decreasing distance between the rods or between rod and channel wall, respectively. To determine the origin of this phenomenon, experiments were performed in rod bundles with different pitch-to-diameter (P/D) and wall-to-diameter (W/D) ratios. In these experiments, two components of the fluctuating velocity were measured with hot wires simultaneously at two different locations of a wall subchannel, together with the pressure fluctuations at the wall measured by microphones. The output signals were registered with an analog tape recorder. Afterwards they were digitized and evaluated to obtain spectra as well as auto and cross correlations. The results were analysed to determine the interdependence between pressure and velocity fluctuations. Attention was devoted to the analysis of turbulence spectra and the identification of their specific ranges. The dominant frequency of the turbulent motion, taken from the spectra, was found to be a function of the gap width and of the flow velocity. The corresponding Strouhal number is a geometrical parameter which can be expressed in terms of P/D and W/D. Based on the observation of transit time between the probes, measured with help of cross correlations, on the form and the presence of peaks on spectra, a phenomenological model was developed, to explain the studied phenomenon. The model describes the formation of large eddies near the gaps and their effect on the fluid motion through rod bundles. The relationship between the mixing process and the studied phenomenon was determined. (orig.) [de

  17. Prediction of velocity distributions in rod bundle axial flow, with a statistical model (K-epsilon) of turbulence

    International Nuclear Information System (INIS)

    Silva Junior, H.C. da.

    1978-12-01

    Reactor fuel elements generally consist of rod bundles with the coolant flowing axially through the region between the rods. The confiability of the thermohydraulic design of such elements is related to a detailed description of the velocity field. A two-equation statistical model (K-epsilon) of turbulence is applied to compute main and secondary flow fields, wall shear stress distributions and friction factors of steady, fully developed turbulent flows, with incompressible, temperature independent fluid flowing axially through triangular or square arrays of rod bundles. The numerical procedure uses the vorticity and the stream function to describe the velocity field. Comparison with experimental and analytical data of several investigators is presented. Results are in good agreement. (Author) [pt

  18. Two-phase flow and cross-mixing measurements in a rod bundle

    International Nuclear Information System (INIS)

    Yloenen, A.; Prasser, H.-M.

    2011-01-01

    The wire-mesh sensor technique has been used for the first time to study two-phase flow and liquid mixing in a rod bundle. A dedicated test facility (SUBFLOW) was constructed at Paul Scherrer Institut (PSI) in a co-operation with the Swiss Federal Institute of Technology (ETH Zurich). Simultaneous injection of salt water as tracer and air bubbles can be used to quantify the enhancement of liquid mixing in two-phase flow when the results are compared with the single-phase mixing experiment with the same test parameters. The second aspect in the current experiments is the two-phase flow in bundle geometry. (author)

  19. Rehme correlation for spacer pressure drop compared to XT-ADS rod bundle simulations and water experiment

    International Nuclear Information System (INIS)

    Batta, A.; Class, A.; Litfin, K.; Wetzel, T.

    2011-01-01

    The Rehme correlation is the most common formula to estimate the pressure drop of spacers in the design phase of new bundle geometries. It is based on considerations of momentum losses and takes into account the obstruction of the flow cross section but it ignores the geometric details of the spacer design. Within the framework of accelerator driven sub-critical reactor systems (ADS), heavy-liquid-metal (HLM) cooled fuel assemblies are considered. At the KArlsruhe Liquid metal LAboratory (KALLA) of the Karlsruhe Institute of Technology a series of experiments to quantify both pressure losses and heat transfer in HLM-cooled rod bundles are performed. The present study compares simulation results obtained with the commercial CFD code Star-CCM to experiments and the Rehme correlation. It can be shown that the Rehme correlation, simulations and experiments all yield similar trends, but quantitative predictions can only be delivered by the CFD which takes into account the full geometric details of the spacer geometry. (orig.)

  20. Thermal behavior simulation of a nuclear fuel rod through an eletrically heated rod

    International Nuclear Information System (INIS)

    Lima, R. de C.F. de.

    1984-01-01

    In thermalhydraulic loops the nuclear industry often uses electrically heated rods to simulate power transients, which occur in nuclear fuel rods. The development and design of a electrically heated rod, by supplying the dimensions and materials which should be used in order to yeld the same temperature and heat flux at the surfaces of the nuclear rod and the electrically heated rod are presented. To a given nuclear transient this equality was obtained by fitting the linear power through the lumped parameters technique. (Author) [pt

  1. Test Facility Construction for Flow Visualization on Mixing Flow inside Subchannels of PWR Rod Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seok; Jeon, Byong-Guk; Youn, Young-Jung; Choi, Hae-Seob; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Flow inside rod bundles has a similarity with flow in porous media. To ensure thermal performance of a nuclear reactor, detailed information of the heat transfer and turbulent mixing flow phenomena taking place within the subchannels is required. The subchannel analysis is one of the key thermal-hydraulic calculations in the safety analysis of the nuclear reactor core. At present, subchannel computer codes are employed to simulate fuel elements of nuclear reactor cores and predict the performance of cores under normal operating and hypothetical accident conditions. The ability of these subchannels codes to predict both the flow and enthalpy distribution in fuel assemblies is very important in the design of nuclear reactors. Recently, according to the modern tend of the safety analysis for the nuclear reactor, a new component scale analysis code, named CUPID, and has been developed in KAERI. The CUPID code is based on a two-fluid and three-field model, and both the open and porous media approaches are incorporated. The PRIUS experiment has addressed many key topics related to flow behaviour in a rod bundle. These issues are related to the flow conditions inside a nuclear fuel element during normal operation of the plant or in accident scenarios. From the second half of 2016, flow visualization will be performed by using a high speed camera and image analysis technique, from which detailed information for the two-dimensional movement of single phase flow is quantified.

  2. CFD simulation and validation of turbulent mixing in a rod bundle with vaned spacer grids based on LDV test

    International Nuclear Information System (INIS)

    Chen Xi; Li Songwei; Li Zhongchun; Du Sijia; Zhang Yu; Peng Huanhuan

    2017-01-01

    Spacer grids with mixing vanes are generally used in fuel assemblies of Pressurized Water Reactor (PWR), because that mixing vanes could enhance the lateral turbulent mixing in subchannels. Thus, heat exchangements are more efficient, and the value of departure from nucleate boiling (DNB) is greatly increased. Actually turbulent mixing is composed of two kinds of flows: swirling flow inside the subchannel and cross flow between subchannels. Swirling flow could induce mixing between hot water near the rod and cold water in the center of the subchannel, and may accelerate deviation of the bubbles from the rod surface. Besides, crossing flow help to mixing water between hot subchannels and cold subchannels, which impact relatively large flow area. As a result, how to accurately capture and how to predict the complicated mixing phenomenon are of great concernments. Recently many experimental studies has been conducted to provide detailed turbulent mixing in rod bundle, among which Laser Doppler Velocimetry method is widely used. With great development of Computational Fluid Dynamics, CFD has been validated as an analysis method for nuclear engineering, especially for single phase calculation. This paper presents the CFD simulation and validation of the turbulent mixing induced by spacer grid with mixing vanes in rod bundles. Experiment data used for validation came from 5 x 5 rod bundle test with LDV technology, which is organized by Science and Technology on Reactor System Design Technology Laboratory. A 5 x 5 rod bundle with two spacer grids were used. Each rod has dimension of 9.5 mm in outer diameter and distance between rods is 12.6 mm. Two axial bulk velocities were conducted at 3.0 m/s for high Reynolds number and 1.0 m/s for low Reynolds number. Working pressure was 1.0 bar, and temperature was about 25degC. Two different distances from the downstream of the mixing spacer grid and one from upstream were acquired. Mean axial velocities and turbulent intensities

  3. Experimental comparison of the optical measurements of a cross-flow in a rod bundle with mixing vanes

    International Nuclear Information System (INIS)

    Chang, Seok Kyu; Choo, Yeon Jun; Kim, Bok Deuk; Song, Chul Hwa

    2008-01-01

    The lateral crossflow on subchannels in a rod bundle array was investigated to understand the flow characteristics related to the mixing vane types on a spacer grid by using the PIV technique. For more measurement resolutions, a 5x5 rod bundle was fabricated a 2.6 times larger than the real rod bundle size in a pressurized water reactor. A rod-embedded optic array was specially designed and used for the illumination of the inner subchannels. The crossflow field in a subchannel was characterized by the type and the arrangement of the mixing vanes. At a near downstream location from the spacer grid (z/D h =1) in the case of the split type, a couple of small vortices were generated diagonally in a subchannel. On the other hand, in the case of the swirl type, there was a large elliptic vortex generated in the center of a subchannel. The measurement results were compared with the experimental results which had been performed with the LDV technique at the same test facility. The magnitudes of the flow velocity and the vorticity in PIV results were less than those in LDV measurement results. It was shown that the instantaneous flow fields in a subchannel frequently have quite different shapes from the averaged one

  4. Experimental investigations on the fluid flow through an asymmetric rod bundle (P/D = 1.148, W/D = 1.045)

    International Nuclear Information System (INIS)

    Rehme, K.

    1983-11-01

    Measurements of the distributions of the mean velocity, the wall shear stresses and the turbulence were performed in a wall subchannel of a rod bundle of four parallel rods arranged asymmetrically in a rectangular channel (P/D = 1.148, W/D = 1.045). The Reynolds number of this investigations was Re = 5.88 x 10 4 . The experimental results show that the momentum transport is highly anisotropic especially in the gaps of the rod bundle. Influences of secondary flow cannot be detected in the distribution of the time-mean velocity. The comparison between experimental wall shear stress distributions and those calculated with the VELASCO-code shows discrepancies both in the gap between the rod and channel walls and in the gap between the rods caused by the high momentum transport between the two subchannels. (orig.) [de

  5. Vibration of fuel bundles

    International Nuclear Information System (INIS)

    Chen, S.S.

    1975-06-01

    Several mathematical models have been proposed for calculating fuel rod responses in axial flows based on a single rod consideration. The spacing between fuel rods in liquid metal fast breeder reactors is small; hence fuel rods will interact with one another due to fluid coupling. The objective of this paper is to study the coupled vibration of fuel bundles. To account for the fluid coupling, a computer code, AMASS, is developed to calculate added mass coefficients for a group of circular cylinders based on the potential flow theory. The equations of motion for rod bundles are then derived including hydrodynamic forces, drag forces, fluid pressure, gravity effect, axial tension, and damping. Based on the equations, a method of analysis is presented to study the free and forced vibrations of rod bundles. Finally, the method is applied to a typical LMFBR fuel bundle consisting of seven rods

  6. Models for the cross flow and the turbulent eddy diffusivity in bundles of rods with helical spacers

    International Nuclear Information System (INIS)

    Fernandez y Fernandez, E.; Carajilescov, P.

    1985-01-01

    The fuel elements of a LMFBR type reactor consist of a bundle of rods wrapped by helical wires that work as spacers. The bundle of rods is surrounded by an hexagonal duct. Models for the channel cross flow and for the turbulent eddy diffusivity were developed. In conjunction with these models, the flow redistribution factors permit to estabish a determinist method to calculate the temperature distribution. The obtained results are compared with experimental data available in the literature and with results given by other codes. Although these codes are based on much more complex models, the comparison was very satisfactory. (Author) [pt

  7. Crossflow between subchannels in a 5 x 5 rod-bundle geometry

    Science.gov (United States)

    Lee, Jungjin; Park, Hyungmin

    2017-11-01

    In the present study, we experimentally investigate the single-phase (water as a working fluid) flow in a vertical 5 x 5 rod-bundle geometry using a particle image velociemtry, especially focusing on the crossflow phenomena between subchannels. This crossflow phenomena is very important in determining the performance and safety of nuclear power plant. To measure the flow behind the rod, it is made of FEP (Fluorinated Ethylene Propylene) to achieve the index matching. The ratio of pitch between rods and rod diameter is 1.4, and the considered Reynolds number based on a hydraulic diameter of a channel and an axial bulk velocity is 10000. Also, the typical grid spacer is installed periodically along the streamwise direction. Depending on the location of subchannel (e.g., distance to the side wall or grid spacer), the flow (turbulence) statistics show large variations that will be discussed in detail. Furthermore, we will suggest a modified crossflow model that can explain the varying crossflow phenomena more clearly. Supported by NRF Grant (NRF-2016M2B2A9A02945068) of the Korean government.

  8. Behavior of instantaneous lateral velocity and flow pulsation in duct flow with cylindrical rod

    International Nuclear Information System (INIS)

    Lee, Chi Young; Shin, Chang Hwan; Park, Ju Yong; Oh, Dong Seok; Chun, Tae Hyun; In, Wang Kee

    2012-01-01

    Recently, KAERI (Korea Atomic Energy Research Institute) has examined and developed a dual cooled annular fuel. Dual cooled annular fuel allows the coolant to flow through the inner channel as well as the outer channel. Due to inner channel, the outer diameter of dual cooled annular fuel (15.9 mm) is larger than that of conventional cylindrical solid fuel (9.5 mm). Hence, dual cooled annular fuel assembly becomes a tight lattice fuel bundle configuration to maintain the same array size and guide tube locations as cylindrical solid fuel assembly. P/Ds (pitch between rods to rod diameter ratio) of dual cooled annular and cylindrical solid fuel assemblies are 1.08 and 1.35, respectively. This difference of P/D could change the behavior of turbulent flow in rod bundle. Our research group has investigated a turbulent flow parallel to the fuel rods using two kinds of simulated 3x3 rod bundles. To measure the turbulent rod bundle flow, PIV (Particle Image Velocimetry) and MIR (Matching Index of Refraction) techniques were used. In a simulated dual cooled annular fuel bundle (i.e., P/D=1.08), the quasi periodic oscillating flow motion in the lateral direction, called the flow pulsation, was observed, which significantly increased the lateral turbulence intensity at the rod gap center. The flow pulsation was visualized and measured clearly and successfully by PIV and MIR techniques. Such a flow motion may have influence on the fluid induced vibration, heat transfer, CHF (Critical Heat Flux), and flow mixing between subchannels in rod bundle flow. On the other hand, in a simulated cylindrical solid fuel bundle (i.e., P/D=1.35), the peak of turbulence intensity at the gap center was not measured due to an irregular motion of the lateral flow. This study implies that the behavior of lateral velocity in rod bundle flow is greatly influenced by the P/D (i.e., gap distance). In this work, the influence of gap distance on behavior of instantaneous lateral velocity and flow

  9. Influence on rewetting temperature and wetting delay during rewetting rod bundle by various radial jet models

    Energy Technology Data Exchange (ETDEWEB)

    Debbarma, Ajoy; Pandey, Krishna Murari [National Institute of Technology, Assam (India). Dept. of Mechanical Engineering

    2016-03-15

    Numerical investigation of the rewetting of single sector fuel assembly of Advanced Heavy Water Reactor (AHWR) has been carried out to exhibit the effect of coolant jet diameters (2, 3 and 4 mm) and jet directions (Model: M, X and X2). The rewetting phenomena with various jet models are compared on the basis of rewetting temperature and wetting delay. Temperature-time curve have been evaluated from rods surfaces at different circumference, radial and axial locations of rod bundle. The cooling curve indicated the presence of vapor in respected location, where it prevents the contact between the firm and fluid phases. The peak wall temperature represents as rewetting temperature. The time period observed between initial to rewetting temperature point is wetting delay. It was noted that as improved in various jet models, rewetting temperature and wetting delay reduced, which referred the coolant stipulation in the rod bundle dominant vapor formation.

  10. Influence on rewetting temperature and wetting delay during rewetting rod bundle by various radial jet models

    International Nuclear Information System (INIS)

    Debbarma, Ajoy; Pandey, Krishna Murari

    2016-01-01

    Numerical investigation of the rewetting of single sector fuel assembly of Advanced Heavy Water Reactor (AHWR) has been carried out to exhibit the effect of coolant jet diameters (2, 3 and 4 mm) and jet directions (Model: M, X and X2). The rewetting phenomena with various jet models are compared on the basis of rewetting temperature and wetting delay. Temperature-time curve have been evaluated from rods surfaces at different circumference, radial and axial locations of rod bundle. The cooling curve indicated the presence of vapor in respected location, where it prevents the contact between the firm and fluid phases. The peak wall temperature represents as rewetting temperature. The time period observed between initial to rewetting temperature point is wetting delay. It was noted that as improved in various jet models, rewetting temperature and wetting delay reduced, which referred the coolant stipulation in the rod bundle dominant vapor formation.

  11. Experimental investigations on the fluid flow through an asymmetric rod bundle (P/D = 1.148, W/D = 1.074)

    International Nuclear Information System (INIS)

    Rehme, K.

    1984-12-01

    Measurements of the distributions of the mean velocity, the wall shear stresses and the turbulence were performed in a wall subchannel of a rod bundle of four prallel rods arranged asymmetrically in a rectangular (P/D = 1.148, W/D = 1.074). The Reynolds number of this investigations was Re = 7.89 x 10 4 . The results obtained by a fully automated rig are compared with those from manual operation. The experimental results show that the momentum transport is highly anisotropyc especially in the gaps of the rod bundle. Influences of secondary flow cannot be detected in the distribution of the time-mean velocity. The comparison between experimental wall shear stress distributions and those calculated with the VELASCO-code shows discrepancies both in the gap between the rod and channel walls and in the gap between the rods caused by the high momentum transport between the two subchannels. (orig.) [de

  12. Numerical Simulation for Frictional Loss and Local Loss of a 5*5 SMART Rod Bundle

    International Nuclear Information System (INIS)

    Park, Jong-Pil; Kim, Seong Jin; Kwon, Hyuk; Seo, Kyong-Won; Hwang, Dae-Hyun

    2014-01-01

    The results showed good agreement with experimental data and/or reasonable values. However, these results were dependent on computational meshes and turbulence models and it still remains important issues in CFD analysis. The aim of present work is to assess the pressure drop in a 5*5 SMART rod bundle using 3D CFD code with various computational meshes and turbulence models. In the present work, 3D CFD code was utilized to investigate pressure drop in a SMART 5*5 rod bundle. The predicted pressure drop was strongly dependent with computational meshes and turbulence models. Based on CFD results in this study, least five of six meshes within the subchannel gap are required to get reliable result which is insensitive to the number of meshes. The friction factor predicted by k - ε model is good agreement with McAdams's correlation while SST model overestimate McAdams's correlation. However, it is difficult to judge performance of turbulence model because of lock of experimental data for a 5*5 SMART bare rod bundle. For nominal condition (Re-194,000) of SMART, SST model predict k-factor of MV and IFM grid as 1.304 and 0.748, respectively. This value is reasonable as compared with designed k-factor, 1.320 and 0.78

  13. Heat transfer in tube bundles of heat exchangers with flow baffles induced forced mixing

    International Nuclear Information System (INIS)

    AbuRomia, M.M.; Chu, A.W.; Cho, S.M.

    1976-01-01

    Thermal analysis of shell-and-tube heat exchangers is being investigated through geometric modeling of the unit configuration in addition to considering the heat transfer processes taking place within the tube bundle. The governing equations that characterize the heat transfer from the shell side fluid to the tube side fluid across the heat transfer tubewalls are indicated. The equations account for the heat transfer due to molecular conduction, turbulent thermal diffusion, and forced fluid mixing among various shell side fluid channels. The analysis, though general in principle, is being applied to the Clinch River Breeder Reactor Plant-Intermediate Heat Exchanger, which utilizes flow baffles appropriately designed for induced forced fluid mixing in the tube bundle. The results of the analysis are presented in terms of the fluid and tube wall temperature distributions of a non-baffled and baffled tube bundle geometry. The former case yields axial flow in the main bundle region while the latter is associated with axial/cross flow in the bundle. The radial components of the axial/cross flow yield the necessary fluid mixing that results in reducing the thermal unbalance among the heat transfer to the allowable limits. The effect of flow maldistribution, present on the tube or shell sides of the heat exchangers, in altering the temperature field of tube bundles is also noted

  14. Nucleate boiling heat transfer on horizontal tubes in bundles

    International Nuclear Information System (INIS)

    Fujital, Y.; Ohta, H.; Hidaka, S.; Nishikawa, K.

    1986-01-01

    In order to clarify the heat transfer mechanisms of the flooded type horizontal tube bundle evaporator, heat transfer characteristics of tube bundles of experimental scale which consist both of smooth and enhanced tubes were investigated in detail. The experiments of saturated nucleate boiling were performed by using Freon 113 under pressures 0.1 to 1 MPa, and the effects of various parameters, for example, bundle arrangement, heat flux, pressure on the characteristics of an individual tube are clarified. Experimental data is reproduced well by a proposed heat transfer model in which convective heat transfer coefficients due to rising bubbles are estimated as a function of their volumetric flow rate

  15. Flow field measurements using LDA and numerical computation for rod bundle of reactor fuel assembly

    International Nuclear Information System (INIS)

    Hu Jun; Zou Zunyu

    1995-02-01

    Local mean velocity and turbulence intensity measurements were obtained with DANTEC 55 X two-dimensional Laser Dopper Anemometry (LDA) for rod bundle of reactor fuel assembly test model which was a 4 x 4 rod bundle. The data were obtained from different experimental cross-sections both upstream and downstream of the model support plate. Measurements performed at test Reynolds numbers of 1.8 x 10 4 ∼3.6 x 10 4 . The results described the local and gross effects of the support plate on upstream and downstream flow distributions. A numerical computation was also given, the experimental results are in good agreement with the numerical one and the others in references. Finally, a few suggestions were proposed for how to use the LDA system well. (11 figs.)

  16. Air velocity profiles near sleeve blockages in an unheated 7 x 7 rod bundle. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Creer, J. M.; Bates, J. M.

    1979-04-01

    Local air velocity measurements were obtained with a laser Doppler anemometer near flow blockages in an unheated 7 x 7 rod bundle. Sleeve blockages were positioned on the center nine rods to create an area reduction of 90% in the center four subchannels of the bundle. Experimental results indicated that severe flow disturbances occurred downstream from the blockage cluster but showed only minor flow disturbances upstream from the blockage. Flow reversals were detected downstream from the blockage and persisted for approximately five subchannel hydraulic diameters. The air velocity profiles were in excellent agreement with water velocity data previously obtained at essentially the same Reynolds number. Subchannel average velocity predictions obtained with the COBRA computer program were in good agreement with subchannel average velocities estimated using the measured local velocity data.

  17. Utilization of the MAT method to analyze the nucleate boiling boundary in rod bundles subchannels

    International Nuclear Information System (INIS)

    Pedron, M.Q.

    1983-01-01

    The digital program PANTERA-1P, a new version of the COBRA-IIIC code, developed at CDTN, is directed to the thermal-hydraulic analysis of water cooled rod bundles and reactor cores, insteady state and transient conditions. Both the new and the old code versions have identical capacities in what concerns evaluation of fluid variables, nevertheless PANTERA-1P has better and faster performance. Improvements introduced in the scheme for solution of the conservation equations have contributed significantly to reduce the computer time, without affecting the accuracy of results. While the momentum equations are solved in COBRA-IIIC for the crossflow distribution, the PANTERA-1P code solves these equations for the pressure distribution by using the MAT method (Modified and Advanced Theta). The calculation of the pressure coefficient matrix has been optimized and simultaneous linear equations are solved optionally by means of the transpose elimination with storage requirements or the successive over-relaxation methods. The program presents others features specially in what concerns the thermal conduction model for fuel rods and the critical heat flux calculations options. A new input/output scheme is provided for optional use of the British or Internacional System of Units. The results of the program are compared to the critical heat flux experimental data and to the results of COBRA-IIIC. Excellent agreement is observed in both cases. (Author) [pt

  18. Benchmark thermal-hydraulic analysis with the Agathe Hex 37-rod bundle

    International Nuclear Information System (INIS)

    Barroyer, P.; Hudina, M.; Huggenberger, M.

    1981-09-01

    Different computer codes are compared, in prediction performance, based on the AGATHE HEX 37-rod bundle experimental results. The compilation of all available calculation results allows a critical assessment of the codes. For the time being, it is concluded which codes are best suited for gas cooled fuel element design purposes. Based on the positive aspects of these cooperative Benchmark exercises, an attempt is made to define a computer code verification procedure. (Auth.)

  19. Experiments and correlations of pressure loss coefficients for hexagonal arranged rod bundles (P/D > 1.02) with helical wire spacers in laminar and turbulent flows

    International Nuclear Information System (INIS)

    Marten, K.; Yonekawa, S.; Hoffmann, H.

    1987-05-01

    Advanced pressurized water reactors as well as sodium cooled fast reactors, in their breeding and absorber elements, use tightly packed rod bundles with hexagonally arranged rods. Helical wires or helical fins serve as spacers. The pressure loss coefficients of twelve bundles with helical wires were determined systematically in water experiments. High measuring accuracy was achieved by very precise fabrication of the bundles and the shroud as well as by investigations of the proper measuring techniques. The results show a dependency of the loss coefficients on the Reynolds number and on the P/D and H/D ratios of the bundles. These results together with available systematic experimental results of investigations at P/D > 1.1 were used to develop a correlation to determine the pressure loss coefficients of tightly and widely packed hexagonally arranged rod bundles with helical wire spacers. These correlations were used to recalculate and compare results of pressure loss investigations found in the literature; good agreement was demonstrated. Hence, calculation methods exist for a broad range of applications to determine the pressure loss coefficients of hexagonally arranged rod bundles with helical wires for spacers. (orig./HP) [de

  20. Experimental study of static pressure distribution and axial pressure drop in a seven wire-wrapped rod bundle

    International Nuclear Information System (INIS)

    Fernandez y Fernandez, E.; Carajilescov, P.

    1980-11-01

    The fuel element of a LMFBR type reactor consists of a rod bundle in a triangular array with helicoidal spacers among which the coolant flows. By utilizing a seven wire-wrapped rod bundle, coupled to an air loop, the hydrodynamic behaviour of the flow was simulated. A series of measurements was performed in order to obtain static pressure distributions in the surface of the rods and in the walls of the hexagonal duct, for different Reynolds numbers, the axial and the angular position being varied. The axial pressure drop was also measured and the friction coefficient for different Reynolds numbers was calculated. From the results obtained, the existence of zones of low pressure on the surface of the rods was observed, as well as the non-dependence of the nondimensional static pressure on the Reynolds number. Sudden variations in the distribution of the static pressure distribution were observed and they must be taken in to account in the thermal-hydraulic design, due to the possibility of occurence of cavitation bubbles in the coolant. (I.C.R.) [pt

  1. Measurement and modeling of two-phase flow parameters in scaled 8 Multiplication-Sign 8 BWR rod bundle

    Energy Technology Data Exchange (ETDEWEB)

    Yang, X.; Schlegel, J.P.; Liu, Y.; Paranjape, S.; Hibiki, T. [School of Nuclear Engineering, Purdue University, 400 Central Dr., West Lafayette, IN 47907-2017 (United States); Ishii, M., E-mail: ishii@purdue.edu [School of Nuclear Engineering, Purdue University, 400 Central Dr., West Lafayette, IN 47907-2017 (United States)

    2012-04-15

    Highlights: Black-Right-Pointing-Pointer Grid spacers have a significant but not well understood effect on flow behavior and development. Black-Right-Pointing-Pointer Two different length scales are present in rod bundles, which must be accounted for in modeling. Black-Right-Pointing-Pointer An easy-to-implement empirical model has been developed for the two-phase friction multiplier. - Abstract: The behavior of reactor systems is predicted using advanced computational codes in order to determine the safety characteristics of the system during various accidents and to determine the performance characteristics of the reactor. These codes generally utilize the two-fluid model for predictions of two-phase flows, as this model is the most accurate and detailed model which is currently practical for predicting large-scale systems. One of the weaknesses of this approach however is the need to develop constitutive models for various quantities. Of specific interest are the models used in the prediction of void fraction and pressure drop across the rod bundle due to their importance in new Natural Circulation Boiling Water Reactor (NCBWR) designs, where these quantities determine the coolant flow rate through the core. To verify the performance of these models and expand the existing experimental database, data has been collected in an 8 Multiplication-Sign 8 rod bundle which is carefully scaled from actual BWR geometry and includes grid spacers to maintain rod spacing. While these spacer grids are 'generic', their inclusion does provide valuable data for analysis of the effect of grid spacers on the flow. In addition to pressure drop measurements the area-averaged void fraction has been measured by impedance void meters and local conductivity probes have been used to measure the local void fraction and interfacial area concentration in the bundle subchannels. Experimental conditions covered a wide range of flow rates and void fractions up to 80%.

  2. Development of drift-flux model based on 8 x 8 BWR rod bundle geometry experiments under prototypic temperature and pressure conditions

    International Nuclear Information System (INIS)

    Ozaki, Tetsuhiro; Suzuki, Riichiro; Mashiko, Hiroyuki; Hibiki, Takashi

    2013-01-01

    The drift-flux model is one of the imperative concepts used to consider the effects of phase coupling on two-phase flow dynamics. Several drift-flux models are available that apply to rod bundle geometries and some of these are implemented in several nuclear safety analysis codes. However, these models are not validated by well-designed prototypic full bundle test data, and therefore, the scalability of these models has not necessarily been verified. The Nuclear Power Engineering Corporation (NUPEC) conducted void fraction measurement tests in Japan with prototypic 8 x 8 BWR (boiling water reactor) rod bundles under prototypic temperature and pressure conditions. Based on these NUPEC data, a new drift-flux model applicable to predicting the void fraction in a rod bundle geometry has been developed. The newly developed drift-flux model is compared with the other existing data such as the two-phase flow test facility (TPTF) data taken at the Japan Atomic Energy Research Institute (JAERI) [currently, Japan Atomic Energy Agency (JAEA)] and low pressure adiabatic 8 x 8 bundle test data taken at Purdue University in the United States. The results of these comparisons show good agreement between the test data and the predictions. The effects of power distribution, spacer grids, and the bundle geometry on the newly developed drift-flux model have been discussed using the NUPEC data. (author)

  3. Generation of heat on fuel rod in cosine pattern by using induction heating

    International Nuclear Information System (INIS)

    Keettikkal, Felix; Sajeesh, Divya; Rao, Poornima; Hande, Shashank; Dakave, Ganesh; Kute, Tushar; Mahajan, Akshay; Kulkarni, R.D.

    2017-01-01

    Fuel rods are used in a nuclear reactor for fission process. When these rods are cooled by water during the heat transfer, the temperature stress causes undesirable defects in the fuel rod. Studying these defects occurring in the fuel rod in the nuclear cluster during nuclear reaction is a difficult task because fission reaction makes it difficult to analyse the changes in the rod. Hence there is a need to use a replica of the rod with similar thermal stress to study and analyse the rod for the defects. Normally the heat generated on the fuel rod follows a cosine pattern which is an inherent characteristic inside a nuclear reactor. In view of this, in this paper induction heating method is used on a rod to create an exact replica of the cosine pattern of heat by varying the pitch of the coil. First, a MATLAB simulation is done using simulink. Then a prototype of the model has been developed comprising of carbon steel pipe, with length and outside diameter of 1 meter and 48.2 mm, respectively. Instead of using water as coolant, rod is simulated in air. Therefore, the heat generated is lost by normal convection and radiation. Non-nuclear testing can be a valuable tool in the development or in some kind of experiment using nuclear reactor. Induction heating becomes an alternative to classical heating technologies because of its advantages such as efficiency, quickness, safety, clean heating and accurate power control. (author)

  4. Subchannel friction factors for rod bundles: laminar flow predictions and their application to turbulent flows

    International Nuclear Information System (INIS)

    Robinson, D.P.

    1979-02-01

    For the calculation of friction factors the use of correlations validated for smooth circular tubes along with the duct hydraulic diameter is known to be inappropriate for certain non-circular geometries. In order to test the validity and range of application of such correlations to the subchannels of rod bundles a computer programme has been written for the prediction of subchannel laminar velocity distributions and friction coefficients for fully developed flow. The theoretical basis and development of the programme is described along with comparisons between predictions and existing solutions for some simple geometries. Using the computer programme a wide range of calculations have been carried out for flow sections representing edge, corner and internal subchannels of rod bundles with particular emphasis on those of in-line pin bundle geometries. Where comparison can be made the predicted laminar coefficients are in excellent agreement with existing solutions. Although the approach adopted here could be used as the basis of a model for the subchannel axial friction factor, careful account should be taken of enhanced turbulent momentum transfer in situations where the flow is not unidirectional. (UK)

  5. Single-phase cross-mixing measurements in a 4 x 4 rod bundle

    International Nuclear Information System (INIS)

    Yloenen, Arto; Bissels, Wilhelm-Martin; Prasser, Horst-Michael

    2011-01-01

    Highlights: → The wire-mesh sensor technique has been successfully introduced into a fuel rod bundle geometry. → Quantitative information on the turbulent dispersion of the fluid was obtained. → In full spatial and temporal resolution, the data is interesting for the unsteady CFD validation. - Abstract: The wire-mesh sensor technique has been successfully introduced into a fuel rod bundle geometry for the first time. In this context, a dedicated test facility (SUBFLOW) has been designed and constructed at Paul Scherrer Institut (PSI) in a co-operation with the Swiss Federal Institute of Technology (ETH Zuerich). Two wire-mesh sensors designed and built in-house were installed in the upper part of the vertical test section of SUBFLOW, and single-phase experiments on the turbulent mass exchange between neighboring sub-channels were performed. For this purpose, salt tracer was injected locally in one of the sub-channels and conductivity distributions in the bundle measured by the wire-mesh sensor. Both flow rate and distance from the injection point were varied. The latter was achieved by using injection nozzles at different heights. In this way, the sensor located in the upper part of the channel could be used to characterize the progress of the mixing along the flow direction, and the degree of cross-mixing assessed using the quantity of tracer arriving in the neighboring sub-channels. Fluctuations of the tracer concentration in time were used for statistical evaluations, such as the calculation of standard deviations and two-point correlations.

  6. Method for calculating the critical heat flux in mixed rod assemblies based on the tables of crisis in bundles

    International Nuclear Information System (INIS)

    Bobkov, V.P.

    2000-01-01

    The method for calculating the critical heat flux in the mixed rod assemblies, for example RBMK, containing three-four angle and peripheral macrocells, is presented. The method is based on generalization of experimental data in form of tables for the rods beams. It is recommended for the areas of parameters both provided for by experimental data and for others, where the data are absent. The advantages of the table method as follows: it is acceptable within a wide range of parameters and provides for smooth description of dependence of critical heat fluxes on these parameters; it is characterized by clearness, high reliability and accuracy and is easy in application [ru

  7. Comparing studies for an optimization of steam-heated tube bundle heat exchangers

    International Nuclear Information System (INIS)

    Horn, M.

    1975-01-01

    The problems of designing an apparatus are to be shown by the example of the steam-heated tube bundle heat exchanger, and optimizations are to be carried through by relevant examples. From the results of the optimization, a set of apparatus types is to be derived where the dimensions of the shell and the heat pipes as well as the length of the tube bundle are to be determined by as few data as possible. (orig./TK) [de

  8. Simulation of nuclear fuel rods by using process computer-controlled power for indirect electrically heated rods

    International Nuclear Information System (INIS)

    Malang, S.

    1975-11-01

    An investigation was carried out to determine how the simulation of nuclear fuel rods with indirect electrically heated rods could be improved by use of a computer to control the electrical power during a loss-of-coolant accident (LOCA). To aid in the experiment, a new version of the HETRAP code was developed which simulates a LOCA with heater rod power controlled by a computer that adjusts rod power during a blowdown to minimize the difference in heat flux of the fuel and heater rods. Results show that without computer control of heater rod power, only the part of a blowdown up to the time when the heat transfer mode changes from nucleate boiling to transition or film boiling can be simulated well and then only for short times. With computer control, the surface heat flux and temperature of an electrically heated rod can be made nearly identical to that of a reactor fuel rod with the same cooling conditions during much of the LOCA. A small process control computer can be used to achieve close simulation of a nuclear fuel rod with an indirect electrically heated rod

  9. Optimization of a fuel bundle within a CANDU supercritical water reactor

    International Nuclear Information System (INIS)

    Schofield, M.E.

    2009-01-01

    The supercritical water reactor is one of six nuclear reactor concepts being studied under the Generation IV International Forum. Generation IV nuclear reactors will improve the metrics of economics, sustainability, safety and reliability, and physical protection and proliferation resistance over current nuclear reactor designs. The supercritical water reactor has specific benefits in the areas of economics, safety and reliability, and physical protection. This work optimizes the fuel composition and bundle geometry to maximize the fuel burnup, and minimize the surface heat flux and the form factor. In optimizing these factors, improvements can be achieved in the areas of economics, safety and reliability of the supercritical water reactor. The WIMS-AECL software was used to model a fuel bundle within a CANDU supercritical water reactor. The Gauss' steepest descent method was used to optimize the above mentioned factors. Initially the fresh fuel composition was optimized within a 43-rod CANFLEX bundle and a 61-rod bundle. In both the 43-rod and 61-rod bundle scenarios an online refuelling scheme and non-refuelling scheme were studied. The geometry of the fuel bundles was then optimized. Finally, a homogeneous mixture of thorium and uranium fuel was studied in a 60-rod bundle. Each optimization process showed definitive improvements in the factors being studied, with the most significant improvement being an increase in the fuel burnup. The 43-rod CANFLEX bundle was the most successful at being optimized. There was little difference in the final fresh fuel content when comparing an online refuelling scheme and non-refuelling scheme. Through each optimization scenario the ratio of the fresh fuel content between the annuli was a significant determining cause in the improvements in the factors being optimized. The geometry optimization showed that improvement in the design of a fuel bundle is indeed possible, although it would be more advantageous to pursue it

  10. CODEX-B4C experiment. Core degradation test with boron carbide control rod

    International Nuclear Information System (INIS)

    Hozer, Z.; Nagy, I.; Windberg, P.; Balasko, M.; Matus, L.; Prokopiev, O.; Pinter, A.; Horvath, M.; Gyenes, Gy.; Czitrovszky, A.; Nagy, A.; Jani, P.

    2003-11-01

    The CODEX-B4C bundle test has been successfully performed on 25 th May 2001 in the framework of the COLOSS project of the EU 5 th FWP. The high temperature degradation of a VVER-1000 type bundle with B 4 C control rod was investigated with electrically heated fuel rods. The experiment was carried out according to a scenario selected in favour of methane formation. Degradation of control rod and fuel bundle took place at temperatures ∼2000 deg C, cooling down of the bundle was performed in steam atmosphere. The gas composition measurement indicated no methane production during the experiment. High release of aerosols was detected in the high temperature oxidation phase. The on-line measured data are collected into a database and are available for code validation and development. (author)

  11. Numerical prediction of pressure loss in tight-lattice rod bundle by use of 3-dimensional two-fluid model simulation code ACE-3D

    International Nuclear Information System (INIS)

    Yoshida, Hiroyuki; Takase, Kazuyuki; Suzuki, Takayuki

    2009-01-01

    Two-fluid model can simulate two-phase flow by computational cost less than detailed two-phase flow simulation method such as interface tracking method or particle interaction method. Therefore, two-fluid model is useful for thermal hydraulic analysis in large-scale domain such as a rod bundle. Japan Atomic Energy Agency (JAEA) develops three dimensional two-fluid model analysis code ACE-3D that adopts boundary fitted coordinate system in order to simulate complex shape flow channel. In this paper, boiling two-phase flow analysis in a tight-lattice rod bundle was performed by the ACE-3D. In the results, the void fraction, which distributes in outermost region of rod bundle, is lower than that in center region of rod bundle. The tendency of void fraction distribution agreed with the measurement results by neutron radiography qualitatively. To evaluate effects of two-phase flow model used in the ACE-3D, numerical simulation of boiling two-phase in tight-lattice rod bundle with no lift force model was also performed. In the results, the lift force model has direct effects on void fraction concentration in gap region, and pressure distribution in horizontal plane induced by void fraction distribution cause of bubble movement from the gap region to the subchannel region. The predicted pressure loss in the section that includes no spacer accorded with experimental results with around 10% of differences. The predicted friction pressure loss was underestimated around 20% of measured values, and the effect of the turbulence model is considered as one of the causes of this underestimation. (author)

  12. Experimental data base of turbulent flow in rod bundles using laser doppler velocimeter

    International Nuclear Information System (INIS)

    Chung, Moon Ki; Yang, Sun Kyu; Chung, Heung June; Won, Soon Yeun; Kim, Bok Deuk; Cho, Young Rho

    1992-01-01

    This report presents in detail the hydraulic characteristics measurements in subchannels of rod bundles using one-component LDV (Laser Doppler Velocimeter). In particular, this report presents the figures and tabulations of the resulting data. The detailed explanations about these results are shown in references publicated or presented at the conference. 4 kinds of experimental work were performed so far. (Author)

  13. Turbulent flow through a wall subchannel of a rod bundle

    International Nuclear Information System (INIS)

    Rehme, K.

    1978-04-01

    The turbulent flow through a wall subchannel of a rod bundle was investigated experimentally by means of hotwires und Pitot-tubes. The aim of this investigation was to get experimental information on the transport properties of turbulent flow especially on the momentum transport. Detailed data were measured of the distributions of the time-mean velocity, the turbulence intensities and, hence the kinetic of turbulence, of the shear stresses in the directions normal and parallel to the walls, and of the wall shear stresses. The pitch-to-diameter ratio of the rods equal to the wall-to-diameter ratio was 1.15, the Reynolds number of this investigation was Re = 1.23.10 5 . On the basis of the measurements the eddy viscosities normal and parallel to the walls were calculated. The eddy viscosities observed showed a considerable deviation from the data known up-to-now and from the assumptions introduced in the codes. (orig.) [de

  14. Application of fast neutron radiography to three-dimensional visualization of steady two-phase flow in a rod bundle

    CERN Document Server

    Takenaka, N; Fujii, T; Mizubata, M; Yoshii, K

    1999-01-01

    Three-dimensional void fraction distribution of air-water two-phase flow in a 4x4 rod-bundle near a spacer was visualized by fast neutron radiography using a CT method. One-dimensional cross sectional averaged void fraction distribution was also calculated. The behaviors of low void fraction (thick water) two-phase flow in the rod bundle around the spacer were clearly visualized. It was shown that the void fraction distributions were visualized with a quality similar to those by thermal neutron radiography for low void fraction two-phase flow which is difficult to visualize by thermal neutron radiography. It is concluded that the fast neutron radiography is efficiently applicable to two-phase flow studies.

  15. CODEX-B4C experiment. Core degradation test with boron carbide control rod

    Energy Technology Data Exchange (ETDEWEB)

    Hozer, Z; Nagy, I; Windberg, P; Balasko, M; Matus, L; Prokopiev, O; Pinter, A; Horvath, M; Gyenes, Gy [KFKI Atomic Energy Research Institute, Budapest (Hungary); Czitrovszky, A; Nagy, A; Jani, P [Research Institute for Solid State Physics and Optics, Budapest (Hungary)

    2003-11-01

    The CODEX-B4C bundle test has been successfully performed on 25{sup th} May 2001 in the framework of the COLOSS project of the EU 5{sup th} FWP. The high temperature degradation of a VVER-1000 type bundle with B{sub 4}C control rod was investigated with electrically heated fuel rods. The experiment was carried out according to a scenario selected in favour of methane formation. Degradation of control rod and fuel bundle took place at temperatures {approx}2000 deg C, cooling down of the bundle was performed in steam atmosphere. The gas composition measurement indicated no methane production during the experiment. High release of aerosols was detected in the high temperature oxidation phase. The on-line measured data are collected into a database and are available for code validation and development. (author)

  16. Experimental investigations on the fluid flow through a wall subchannel of a rod bundle (P/D = 1.036, W/D = 1.072)

    International Nuclear Information System (INIS)

    Rehme, K.

    1982-07-01

    Measurements of the distributions of the mean velocity, the wall shear stresses and the turbulence were performed in a wall subchannel of a rod bundle of four parallel rods arranged symmetrically in a rectangular channel (P/D = 1.036, W/D = 1.072). The Reynolds number of this investigation was Re = 7.60 x 10 4 . The experimental results show that the momentum transport is highly anisotropic especially in the gaps of the rod bundle. Influences of secondary flow cannot be detected in the distribution of the time-mean velocity, however, such influences are found in the distributions of the turbulence intensities and the kinetic energy of turbulence. Very high turbulence intensities were observed in the gap between the rods. The comparison between experimental wall shear stress distributions and those calculated with the VELASCO-code shows discrepancies especially in the gap between the rods. (orig.) [de

  17. Interactions in Zircaloy/UO2 fuel rod bundles with Inconel spacers at temperatures above 1200deg C (posttest results of severe fuel damage experiments CORA-2 and CORA-3)

    International Nuclear Information System (INIS)

    Hagen, S.; Hofmann, P.; Schanz, G.; Sepold, L.

    1990-09-01

    In the CORA experiments test bundles of usually 16 electrically heated fuel rod simulators and nine unheated rods are subjected to temperature transients of a slow heatup rate in a steam environment. Thus, an accident sequence is simulated, which may develop from a small-break loss-of-coolant accident of an LWR. An aim of CORA-2, as a first test of its kind, was also to gain experience in the test conduct and posttest handling of UO 2 specimens. CORA-3 was performed as a high-temperature test. The transient phases of CORA-2 and CORA-3 were initiated with a temperature ramp rate of 1 K/s. The temperature escalation due to the exothermal zircaloy(Zry)-steam reaction started at about 1000deg C, leading the bundles to maximum temperatures of 2000deg C and 2400deg C for tests CORA-2 and CORA-3, respectively. The test bundles resulted in severe oxidation and partial melting of the cladding, fuel dissolution by Zry/UO 2 interaction, complete Inconel spacer destruction, and relocation of melts and fragments to lower elevations in the bundle, where extended blockages have formed. In both tests the fuel rod destruction set in together with the formation of initial melts from the Inconel/Zry interaction. The lower Zry spacer acted as a catcher for relocated material. In test CORA-2 the UO 2 pellets partially disintegrated into fine particles. This powdering occurred during cooldown. There was no physical disintegration of fuel in test CORA-3. (orig./MM) [de

  18. Multi-rod burst test under a loss-of coolant accident condition, (4)

    International Nuclear Information System (INIS)

    Otomo, Takashi; Hashimoto, Masao; Kawasaki, Satoru; Furuta, Teruo; Uetsuka, Hiroshi

    1983-06-01

    Multi-rod burst test of No.7808 bundle was performed in steam to estimate quantitative coolant flow channel restriction caused by the ballooning of zircaloy claddings in a fuel assembly during a LOCA transient in LWRs. The test was conducted under the condition that the initial internal pressure in each rod was 35kg/cm 2 (RT) and the heating rate was 9 0 C/s in steam with flow rate of 0.4g/cm 2 .min. The following results were obtained; (1) Maximum and burst pressures in rods were in the range 45 to 48kg/cm 2 and 41 to 45kg/cm 2 , respectively. The burst temperature of cladding were estimated to be 850 to 880 0 C. (2) Axial portions of tubes with greater than 34% strain were observed in the range 0 to 40mm in most rod. The mean length was 19mm in the bundle. (3) The degree of maximum increase in cross-sectional area is 54.2% in the bundle(7 x 7) and 66.9% in the internal rods(5 x 5). (4) Maximum channel area restriction was 40.5% in the bundle(7 x 7) and 51.4% in the internal rods(5 x 5). (author)

  19. Numerical investigation of heat transfer characteristic of fixed planar elastic tube bundles

    International Nuclear Information System (INIS)

    Duan, Derong; Ge, Peiqi; Bi, Wenbo

    2015-01-01

    Highlights: • Both tube-side and shell-side of planar elastic tube bundles were investigated. • Heat transfer and fluid flow were studied from the local analysis perspective. • Secondary flow varies depending on the fluid flow state and the geometry of tube. • Curvature plays a role on the external flow field. • The heat transfer of the two intermediate tube bundles is augmented. - Abstract: Planar elastic tube bundles are a novel approach to enhance heat transfer by using flow-induced vibration. This paper studied the heat transfer characteristic and fluid flow in both tube-side and shell-side using numerical simulation. Two temperature difference formulas were used to calculate convective heat transfer coefficient and the results were verified by theoretical analysis and experimental correlations. The effect of Reynolds number on overall convective heat transfer coefficient and pressure drop in tube-side and shell-side were studied. The comparison of the secondary flow in planar elastic tube bundles and conical spiral tube bundles were conducted. The external flow field and local convective heat transfer around the periphery of fixed planar elastic tube bundles subjected to the cross fluid flow were also analyzed. The results show that the energy consumption efficiency should be taken into account in the forced heat transfer process conducted by adjusting the fluid flow. The secondary flow varies depending on the fluid flow state and the geometry of tube. Hence, it is deduced that the heat transfer enhancement is obtained because the thermal boundary layer in the deformed planar elastic tube bundles caused by flow-induced vibration is damaged by the disordered secondary flow. In addition, the convective heat transfer capability of outside the two intermediate tube bundles is enhanced because of the effect of irregular and complex fluid flow affected by the role of curved tubes on both sides

  20. Use of the ''Lagrangian and Eulerian points of view'' in the transient critical heat flux calculations for BWR rod bundles and experimental verifications

    International Nuclear Information System (INIS)

    Marinelli, V.; Pellei, A.; Vallero, P.; Vitanza, C.

    1975-01-01

    The calculations performed in comparison of the ''Lagrangian point of view'', by means of the DOLCE computer code with the local space--time approach of the ''Eulerian point of view'' indicate that the two methods give substantially equivalent results and predict satisfactorily the onset of the transient CHF for the Centro Informazioni Studi Esperienze annuli experimental data and General Electric Company 16-rod bundles data under typical boiling water reactor transients, including loss-of-coolant accident simulations. 9 references

  1. Steady state heat transfer of helium cooled cable bundles

    International Nuclear Information System (INIS)

    Khalil, A.

    1982-01-01

    In the present study nucleate and film boiling heat transfer characteristics of horizontal conductor bundles are investigated at steady state conditions. The effect of gaps between wires, number of wires, wire position, wire size and bundle orientation on the departure from nucleate boiling and transition to film boiling is studied. For gaps close to the bubble departure diameter, the critical heat flux can approach up to 90% of the single wire value. Consequently, the maximum stable current for a given bundle can be significantly increased above the single conductor value for the same cross-sectional area. (author)

  2. Heat split imbalance study for annular fuel rod

    International Nuclear Information System (INIS)

    He Xiaojun; Ji Songtao; Zhang Yingchao

    2014-01-01

    Annular fuel rod has two gaps at inner and outer side. Under irradiation condition, the dimensional change of pellets is always larger than claddings' due to thermal expansion, swelling and densification, and this tends to enlarge the inner gap and reduce the outer gap. The gap size asymmetry must induce heat split imbalance problem that the heat flux will be larger at outer side of the rod. In this work, computer code AFPAC l.0 is used to simulate this heat split imbalance phenomena. The effect of initial gap size, rod inner pressure, roughness of pellets and cladding is studied, the results reveal that: l) Adjusting initial size of both gaps, reducing inner gap and enlarging outer gap could effectively alleviate heat split imbalance problem; 2) Adjusting the initial roughness of pellets and cladding is another effective approach to reducing heat split imbalance; 3) It seems that changing the rod inner pressure has a little effect on solving the heat flux asymmetry problem. (authors)

  3. Comparison of ASSERT subchannel code with Marviken bundle data

    International Nuclear Information System (INIS)

    Tahir, A.; Carver, M.B.

    1984-04-01

    In this paper ASSERT predictions are compared with the Marviken 6-rod bundle and 36+1 rod bundle. The predictions are presented for two experiments in the 6-rod bundle and four experiments in the 36+1 rod bundle. For low inlet subcooling, the void predictions are in good agreement with the experimental data. For high inlet subcooling, however, the agreement is not as good. This is attributed to the fact that in the high inlet subcooling experiments, single phase turbulent mixing plays a more important role in determining flow conditions in the bundle

  4. Numerical simulation of flow behavior in tight lattice rod bundle

    International Nuclear Information System (INIS)

    Yu Yiqi; Yang Yanhua; Gu Hanyang; Cheng Xu; Song Xiaoming; Wang Xiaojun

    2009-01-01

    The Numerical investigation is performed on the air turbulent flow in triangular rod bundle array. Based on the experimental data, the eddy viscosity turbulent model and the Reynold stress turbulent model are evaluated to simulate the flow behavior in the tight lattice. The results show that SSG Reynolds Stress Model has shown superior predictive performance than other Reynolds-stress models, which indicates that the simulation of the anisotropy of the turbulence is significant in the tight lattice. The result with different Reynolds number and geometry shows that the magnitude of the secondary flow is almost independent of the Reynolds number, but it increases with the decrease of the P/D. (authors)

  5. Experimental study of heavy-liquid metal (LBE) flow and heat transfer along a hexagonal 19-rod bundle with wire spacers

    Energy Technology Data Exchange (ETDEWEB)

    Pacio, J., E-mail: julio.pacio@kit.edu; Daubner, M.; Fellmoser, F.; Litfin, K.; Wetzel, Th.

    2016-05-15

    Highlights: • A unique experiment with lead–bismuth eutectic (LBE) as working fluid was performed. • Detailed temperature measurements were implemented at three axial positions. • The experimental results present a good repeatability within the uncertainties. • Pressure drop results agree with water correlations, as expected. • The Nusselt number is well predicted by the most conservative correlation. - Abstract: An experimental campaign considering a 19-pin hexagonal rod bundle with wire spacers, cooled by forced-convective LBE was completed at the Karlsruhe Liquid Metal Laboratory (KALLA). In the frame of the European research project SEARCH (Safe Exploitation Related Chemistry for HLM Reactors, 2011–2015) the geometry and operating conditions of temperature, flow velocity and power density are representative of the fuel assemblies envisaged for the MYRRHA reactor. An extensive test matrix is evaluated, with 33 experimental runs covering a wide range of Reynolds (ca. 14 000–48 000) and Péclet (ca. 400–1500) numbers, as well as thermal powers (up to 295 kW) at 200 °C inlet temperature, indicating a good degree of reproducibility within the relatively small experimental uncertainties. Both the pressure drop and heat transfer performances are studied. When possible, a comparison with correlations available in the reviewed literature (namely, friction and heat transfer coefficients) is given. Furthermore, the detailed cross-sectional temperature distribution at three selected axial positions is obtained in the experiments and represents the main validation data for CFD. In non-dimensional terms, these profiles could be repeated at different operating conditions, for example hot and cold spots are consistently found at given locations.

  6. Steady-state, local temperature fields with turbulent sodium flow in nominal and disturbed bundle geometries with spacer grids

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.; Kolodziej, M.

    1980-12-01

    The operating reliability of nuclear reactors calls for a reliable strength analysis of the highly loaded core elements, one of its prerequisites being the reliable determination of the three-dimensional velocity and temperature fields. To verify thermohydraulics computer programs, extensive local temperature measurements in the rod claddings of the critical bundle zone were performed on a heated 19-rod bundle model with sodium flow and provided with spacer grids (P/D = 1.30; W/D = 1.19). These are the essential results obtained: Outside the spacer grids the azimuthal temperature variations of the side and corner rods are greater by approximately the factor 10 in the bundle geometry under consideration as compared to rods in the central bundle zone. The spacer grids investigated give rise to great local temperature peaks and correspondingly great temperature gradients in the axial and azimuthal directions immediately around the support points. Continuous reduction of a subchannel by rod bowing results in substantial rises of temperature which, however, are limited to the adjacent cladding tube zones. (orig.) [de

  7. Turbulent interchange in simulated rod bundle geometries for Genetron-12 flows

    International Nuclear Information System (INIS)

    Petrunik, K.

    1973-01-01

    Turbulent interchange data between subchannel arrays simulating an infinite triangular array in a rod bundle fuel cluster were obtained for two-phase Genetron-12 (dichlorodifluoromethane), single phase subcooled Genetron-12 and single phase water flows at gap spacings of 0.025, 0.052 and 0.100 inches. Single phase turbulent interchange rates were relatively independent of the pitch to diameter ratio for the larger two gaps studied but increased for the smallest gap spacing. Two-phase Genetron-12 interchange data were obtained under conditions of unequal qualities and mass fluxes and essentially zero radial pressure gradient along the interconnection region between subchannels. Vapour transport occurred primarily by a diffusional type mechanism and was qualitatively similar to single phase behaviour. For annular flow conditions liquid interchange occurred through a dual mechanism via the film flow and entrained droplets. Vapour interchange was significantly suppressed at the smallest gap spacing due to the presence of the liquid film. Liquid interchange under two-phase conditions increased with gap spacing from 0.025 to 0.052 inches and levelled off slightly at 0.100 inches. Data obtained with heat addition in one test channel indicated negligible effects on the vapour transfer rates but a slight reduction in the magnitude of liquid interchange. (O.T.)

  8. Advances of study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Akira Ohnuki; Kazuyuki Takase; Masatoshi Kureta; Hiroyuki Yoshida; Hidesada Tamai; Wei Liu; Toru Nakatsuka; Hajime Akimoto

    2005-01-01

    R and D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) is started at Japan Atomic Energy Research Institute in collaboration with power company, reactor vendors, universities since 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured LWR technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important R and D items for the RMWR because of the tight-lattice configuration. In this paper, we will show the R and D plan and describe some advances on experimental and analytical studies. The experimental study is performed mainly using large-scale (37-rod bundle) test facility and the analytical one aims to develop a predictable technology for geometry effects such as gap between rods, grid spacer configuration etc. using advanced 3-D two-phase flow simulation methods. Steady-state and transient critical power experiments are conducted with the test facility (Gap width between rods: 1.0 mm) and the experimental data reveal the feasibility of RMWR. (authors)

  9. Heat transfer in tube bundles subjected to blockages. Pt. 1

    International Nuclear Information System (INIS)

    Khattab, M.; Mariy, A.; Habib, M.

    1983-01-01

    The present work is carried out on unblocked test section bundle, half blocked, single ballooning and four ballooning blockages. The hydro-thermal performance of the bundle, (4x4) stainless steel, under each of the previous cases are studied. It is found that the existance of blockages increases the eddies and swirling flow streams. Furthermore, the average heat transfer in a bundle without blockages is superior than that with blockages. The percentage decrease of the average heat transfer coefficient with blockages depends on the position and shape of the blockage. Correlations describing average heat transfer, pressure drop and friction factor are established. All experimental tests are carried out under non-boiling region. (orig.) [de

  10. Numerical Investigation of Cross Flow Phenomena in a Tight-Lattice Rod Bundle Using Advanced Interface Tracking Method

    Science.gov (United States)

    Zhang, Weizhong; Yoshida, Hiroyuki; Ose, Yasuo; Ohnuki, Akira; Akimoto, Hajime; Hotta, Akitoshi; Fujimura, Ken

    In relation to the design of an innovative FLexible-fuel-cycle Water Reactor (FLWR), investigation of thermal-hydraulic performance in tight-lattice rod bundles of the FLWR is being carried out at Japan Atomic Energy Agency (JAEA). The FLWR core adopts a tight triangular lattice arrangement with about 1 mm gap clearance between adjacent fuel rods. In view of importance of accurate prediction of cross flow between subchannels in the evaluation of the boiling transition (BT) in the FLWR core, this study presents a statistical evaluation of numerical simulation results obtained by a detailed two-phase flow simulation code, TPFIT, which employs an advanced interface tracking method. In order to clarify mechanisms of cross flow in such tight lattice rod bundles, the TPFIT is applied to simulate water-steam two-phase flow in two modeled subchannels. Attention is focused on instantaneous fluctuation characteristics of cross flow. With the calculation of correlation coefficients between differential pressure and gas/liquid mixing coefficients, time scales of cross flow are evaluated, and effects of mixing section length, flow pattern and gap spacing on correlation coefficients are investigated. Differences in mechanism between gas and liquid cross flows are pointed out.

  11. CHF experiments of tight pitch lattice rod bundles under PWR pressure condition for development of reduced moderation water reactor

    International Nuclear Information System (INIS)

    Araya, Fumimasa; Nakatsuka, Toru; Yoritsune, Tsutomu

    2002-10-01

    In order to improve plutonium utilization, design studies of reduced moderation water reactors which have hard neutron energy spectrum have been carried out at Division of Energy System Research of Japan Atomic Energy Research Institute (JAERI). At present, triangle, tight pitch lattice cores with about 1 mm gap width between fuel rods have been focused in the neutronic core design. Since a degradation of the heat removal from the fuel rods is worried, an evaluation of heat removal capability i.e. critical heat flux becomes one of important evaluation items in the feasibility study. However, any of published data base, which can be applicable to the evaluation on such narrow gap width cores, does not exist. Therefore, in the present study, in order to accumulate applicable data and to confirm applicability of an evaluation methodology of critical heat flux, basic experiments on the critical heat flux were performed using the test sections consisted of 7 heater rods bundles with the gap widths of 1.5, 1.0 and 0.6 mm under the PWR pressure conditions. The present report describes the experimental apparatus, experimental conditions and accumulated data. Analysis results of the data and the applicability of the evaluation methodology used for the design work are also discussed in this report. As the results of the experiment, it was found that the critical heat flux increased as the mass flux and the inlet subcooling increased. In the region of the mass flux less than about 2,000 kg/m 2 /s, the critical heat flux decreased as the gap width decreased. In the larger mass flux region, obvious trend of effects of the gap width on critical heat flux were not observed due to data scatterings. The flow-area-averaged thermal-equilibrium quality at the CHF position was in the higher ranges from 0.3 to 0.8 in the cases of gap widths of 1.0 and 0.6 mm, and 0.1 to 0.3 in the 1.5 mm case. Based on the experimental results such that the CHFs occurred in the higher quality range and

  12. General correlation for prediction of critical heat flux ratio in water cooled channels

    Energy Technology Data Exchange (ETDEWEB)

    Pernica, R.; Cizek, J.

    1995-09-01

    The paper present the general empirical Critical Heat Flux Ration (CHFR) correlation which is valid for vertical water upflow through tubes, internally heated concentric annuli and rod bundles geometries with both wide and very tight square and triangular rods lattices. The proposed general PG correlation directly predicts the CHFR, it comprises axial and radial non-uniform heating, and is valid in a wider range of thermal hydraulic conditions than previously published critical heat flux correlations. The PG correlation has been developed using the critical heat flux Czech data bank which includes more than 9500 experimental data on tubes, 7600 data on rod bundles and 713 data on internally heated concentric annuli. Accuracy of the CHFR prediction, statistically assessed by the constant dryout conditions approach, is characterized by the mean value nearing 1.00 and the standard deviation less than 0.06. Moverover, a subchannel form of the PG correlations is statistically verified on Westinghouse and Combustion Engineering rod bundle data bases, i.e. more than 7000 experimental CHF points of Columbia University data bank were used.

  13. Effect of a blockage length on the coolability during reflood in a 2 × 2 rod bundle with a 90% partially blocked region

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kihwan, E-mail: kihwankim@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Kim, Byung-Jae, E-mail: byoungjae@kaeri.re.kr [School of Mechanical Engineering, Chungnam National University, 99 Daehak-ro, Yuseoung-Gu, Daejeon 34134 (Korea, Republic of); Choi, Hae-Seob, E-mail: hschoi@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Moon, Sang-Ki, E-mail: skmoon@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of); Song, Chul-Hwa, E-mail: chsong@kaeri.re.kr [Korea Atomic Energy Research Institute, Daeduk-daero 989-111, Yuseong-Gu, Daejeon 34057 (Korea, Republic of)

    2017-02-15

    Highlights: • This test was conducted to understand the effect of blockage length on the coolability. • Reflood tests were conducted with blockage simulators for various reflood rates. • The coolability in the downstream of the blockage region is significantly enhanced. - Abstract: If fuel rods are ballooned or rearranged during the reflood phase of a large break loss-of-coolant accident (LBLOCA) in a pressurized-water reactor (PWR), the transient heat transfer behavior is entirely different with those of the intact fuel rods owing to the deformed blockage region. The coolability in the blocked region depends on a complex two-phase heat transfer with various thermal hydraulic conditions. In addition, the blockage characteristics, such as the blockage ratio, length, shape, and configurations, are also significant factors affecting the coolability. In the present study, reflood experiments were carried out to understand the effect of the blockage length upon the coolability by varying the reflooding rates. The experiments were performed in electrically heated 2 × 2 rod bundles with blockage simulators having the same blockage ratio but different blockage lengths. The characteristics of quenching and heat transfer were evaluated to investigate the influence of the blockage region on the coolability. The droplet behaviors were also observed by measuring the droplets velocity and size near the blockage region. The coolability in the downstream region of the blockage was significantly enhanced, owing to the reduced flow area of the sub-channel, intensification of turbulence, and the entrained droplets in the blockage region.

  14. Development of advanced BWR fuel bundle with spectral shift rod (3) -transient analysis of ABWR core with SSR

    International Nuclear Information System (INIS)

    Ikegawa, T.; Chaki, M.; Ohga, Y.; Abe, M.

    2010-01-01

    The spectral shift rod (SSR) is a new type of water rod, utilized instead of the conventional water rod, in which a water level develops during core operation. The water level can be changed according to the fuel channel flow rate. In this study, ABWR plant performance with SSR fuel bundles under transient conditions has been evaluated using the TRACG code. The TRACG code, which can treat three-dimensional hydrodynamic calculations in a reactor pressure vessel, is well suited for evaluating the reactor transient performance with the SSR fuel bundles because it can calculate the water levels in the SSR at each channel grouping and therefore evaluate the core reactivity according to the water level changes in the SSR. 'Generator load rejection with total turbine bypass failure' and 'Recirculation flow control failure with increasing flow' were selected as cases which may increase the reactivity with the increasing water level in the SSR. It was found that the absolute value of the void reactivity coefficient in the SSR core was larger than that in the conventional water rod core because the core averaged void fraction in the SSR core, which has the vapor region above the water level in the SSR, was larger than that in the conventional water rod core. Therefore, AMCPR for the SSR core was a little larger than that for the conventional water rod core; however, the difference was smaller than 0.02 because the inlet of the SSR ascending path was designed to be small enough to prevent the rapid water level increase in the SSR. (authors)

  15. Computation of turbulent flow and heat transfer in subassemblies

    International Nuclear Information System (INIS)

    Slagter, W.

    1979-01-01

    This research is carried out in order to provide information on the thermohydraulic behaviour of fast reactor subassemblies. The research work involves the development of versatile computation methods and the evaluation of combined theoretical and experimental work on fluid flow and heat transfer in fuel rod bundles. The computation method described here rests on the application of the distributed parameter approach. The conditions considered cover steady, turbulent flow and heat transfer of incompressible fluids in bundles of bare rods. Throughout 1978 main efforts were given to the development of the VITESSE program and to the validation of the hydrodynamic part of the code. In its present version the VITESSE program is applicable to predict the fully developed turbulent flow and heat transfer in the subchannels of a bundle with bare rods. In this paper the main features of the code are described as well as the present status of development

  16. Co-planar deformation and thermal propagation behavior in a bundle burst test

    International Nuclear Information System (INIS)

    Uetsuka, Hiroshi; Koizumi, Yasuo; Kawasaki, Satoru

    1980-07-01

    The probability of the suggested feedback mechanism which could lead to co-planar deformation in a bundle burst test was assessed by the data of test and the calculation based on simplified model. Following four points were evaluated. (1) The probability of local deformation during early heat up stage. (2) The relation between the characteristic of heater and the feedback mechanism. (3) Thermal propagation behavior between two adjacent rods during heat up stage. (4) The propagation of ballooning in a bundle. The probability of suggested feedback mechanism was denied in all the evaluation. The feedback mechanism suggested by Burman could not be a controlling mechanism in co-planar deformation in a bundle burst test. (author)

  17. Burnout and intensification of heat transfer in a four-red bundle in freon-12 and water flows

    International Nuclear Information System (INIS)

    Perepelitsa, N.I.; Pomet'ko, R.S.; Sapankevich, A.P.

    1979-01-01

    Results of experimental investigation of burnout and intensification of heat transfer in a four-rod bundle in freon-12 and water flows are presented. Experiments are carried out at mass rates of 1000-2500 kg/(m 2 xc), pressures of 6.86-12.74 MPa, coolant temperatures at the channel entry of 240-300 deg C for water and pressures of 1.06-2.46 MPa, temperatures of 20-65 deg C for freon. Pressure at the channel exit, coolant expense and temperature at the channel entry were kept constant during experiments. Burnout was attained by smooth increase of electric power and was fixed by indexes of thermal pairs placed in tubes. The comparison of experimental results for freon-12 and water showed that regularities of burnout arising and dependence of critical heat flux (critical power) on regime parameters are qualitatively similar. The experiments showed that dynamics of burnout development for bundles with standard lattices and with lattice-turbulators is essentially different. In the first case burnout starting was accompanied by sharp change of heat transfer surface temperature. In the second case under similar conditions surface temperature increased smoothly in strict correspondance with conducted power. Data evaluation coefficient of freon-12 burnout is specified to water

  18. Feasibility study on thermal-hydraulic performance in tight-lattice rod bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Ohnuki, A.; Kureta, M.; Liu, W.; Tamai, H.; Akimoto, H.

    2004-01-01

    Research and development project for investigating thermal-hydraulic performance in tight-lattice rod bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in 2002. The RMWR can attain the favorable characteristics such as effective utilization of uranium resources, multiple recycling of plutonium, high burn-up and long operation cycle, based on matured light-water reactor technologies. MOX fuel assemblies with tight lattice arrangement are used to increase the conversion ratio by reducing the moderation of neutron. Increasing the in-core void fraction also contributes to the reduction of neutron moderation. The confirmation of thermal-hydraulic feasibility is one of the most important issues for the RMWR because of the tight-lattice configuration. The project has mainly consisted of a large-scale thermal-hydraulic test and development of analytical methods named modeling engineering. In the large-scale test, 37-rod bundle experiments can be performed. Steady-state critical power experiments have been achieved in the test facility and the experimental data reveal the feasibility of RMWR

  19. Effect of orientation on critical heat flux in a 3-rod bundle cooled by Freon-12

    International Nuclear Information System (INIS)

    Dimmick, G.R.

    1979-06-01

    Critical heat flux measurements have been made in a segmented 3-rod test section cooled by Freon-12. Three test section orientations were used: vertical, inclined at 11 deg to the vertical, and horizontal. It was found that at flows of less than 2.5 Mg.m -2 .s -1 the transverse gravity force on the inclined and horizontal orientations reduced the magnitude of the critical heat flux and also changed the location of initial dryout when compared to the vertical data. To account for the effect of orientation during correlation of the data, the Reynolds number was modified to include a transverse gravity term. The minimum standard deviation for the data from the three orientations combined was 3.4 percent and less than 3.7 percent for the three orientations separately. (author)

  20. Nuclear fuel bundle disassembly and assembly tool

    International Nuclear Information System (INIS)

    Yates, J.; Long, J.W.

    1975-01-01

    A nuclear power reactor fuel bundle is described which has a plurality of tubular fuel rods disposed in parallel array between two transverse tie plates. It is secured against disassembly by one or more locking forks which engage slots in tie rods which position the transverse plates. Springs mounted on the fuel and tie rods are compressed when the bundle is assembled thereby maintaining a continual pressure against the locking forks. Force applied in opposition to the springs permits withdrawal of the locking forks so that one tie plate may be removed, giving access to the fuel rods. An assembly and disassembly tool facilitates removal of the locking forks when the bundle is to be disassembled and the placing of the forks during assembly of the bundle. (U.S.)

  1. Steady-state, local temperature fields with turbulent liquid sodium flow in nominal and disturbed bundle geometries with spacer grids

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.

    1980-01-01

    The operating reliability of nuclear reactors calls for a reliable strength analysis of the highly loaded core elements, one of its prerequisites being the reliable determination of the three-dimensional velocity and temperature fields. To verify thermohydraulics computer programs, extensive local temperature measurements in the rod claddings of the critical bundle zone were performed on a heated 19-rod bundle model with sodium flow and provided with spacer grids (P/D = 1.30; W/D = 1.19). The essential results are: - Outside the spacer grids, the azimuthal temperature variations of the side and corner rods are approximately 10-fold those of rods in the central bundle zone. - The spacer grids investigated give rise to great local temperature peaks and correspondingly great temperature gradients in the axial and azimuthal directions immediately around the support points. - Continuous reduction of a subchannel by rod bowing results in substantial rises of temperature which, however, are limited to adjacent cladding tubes. (orig.)

  2. The prediction of burnout in non-uniformly heated rod clusters from burnout data for uniformly heated round tubes

    International Nuclear Information System (INIS)

    Barnett, P.G.

    1964-11-01

    The practice of using burnout data for uniformly heated round tubes to predict burnout in non-uniformly heated reactor channels having complex cross sections is examined. At least two hypotheses are involved: (i) a relationship exists between uniform and non-uniform heat flux distributions and (ii) a relationship exists between simple and complex channel cross sections. Use of two such hypotheses each accurate to ± 15% and a correlation of uniformly heated round tube data having an R.M.S. error of 5%, could yield errors of ± 40% in any predicted value; this figure of ± 40% is regarded as a realistic upper limit for design purposes. It is shown that no method can exist for relating different channel cross sections to within ± 15% and existing methods for relating different heat flux distributions incur some errors exceeding 20%. Furthermore, any suggested method for relating different heat flux distributions can be adequately checked only when a sufficiently large number of results are available and then the preferable alternative of correlating the data is possible. It is concluded that no reliable method can exist for predicting burnout in rod bundles from uniformly heated round tube data with sufficient accuracy for design purposes. (author)

  3. Nanofluid Applied Numerical Analysis of Subchannel in Square Rod Bundle for Fusion-Fission Hybrid System

    International Nuclear Information System (INIS)

    Shamim, Jubair Ahmed; Bhowmik, Palash Kumar; Suh, Kune Y.

    2014-01-01

    Most of the traditional ways available in the literature to enhance heat transfer are mainly based on variation of structures like addition of heat surface area such as fins, vibration of heated surface, injection or suction of fluids, applying electrical or magnetic fields, and so forth. Application of these mechanical techniques to a fuel rod bundle will involve not only designing complex geometries but also using many additional mechanisms inside a nuclear reactor core which in turn will certainly increase the manufacturing cost as well as may hamper various safety features essential for sound and uninterrupted operation of a nuclear power reactor. On the other hand, traditional heat transfer fluids such as water, ethylene glycol and oils have inherently low thermal conductivity relative to metals and even metal oxides. In this study the coolant with suspended nano-sized particles in the base fluid is proposed as an alternative to increase heat transfer but minimize flow resistance inside a nuclear reactor core. Due to technical complexities most of the previous studies carried out on heat transfer of suspension of metal oxides in fluids were limited to suspensions with millimeter or micron-sized particles. Such outsized particles may lead to severe problems in heat transfer equipment including increased pressure drop and corrosion and erosion of components and pipe lines. Dramatic advancement in modern science has made it possible to produce ultrafine metallic or nonmetallic particles of nanometer dimension, which has brought a revolutionary change in the research of heat transfer enhancement methods. Due to very tiny particle size and their small volume fraction, problems such as clogging and increased pressure drop are insignificant for nanofluids. Moreover, the relatively large surface area of nanoparticles augments the stability of nanofluid solution and prevents the sedimentation of nanoparticles. Xuan and Roetzel considered two approaches to illustrate

  4. Nanofluid Applied Numerical Analysis of Subchannel in Square Rod Bundle for Fusion-Fission Hybrid System

    Energy Technology Data Exchange (ETDEWEB)

    Shamim, Jubair Ahmed; Bhowmik, Palash Kumar; Suh, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-05-15

    Most of the traditional ways available in the literature to enhance heat transfer are mainly based on variation of structures like addition of heat surface area such as fins, vibration of heated surface, injection or suction of fluids, applying electrical or magnetic fields, and so forth. Application of these mechanical techniques to a fuel rod bundle will involve not only designing complex geometries but also using many additional mechanisms inside a nuclear reactor core which in turn will certainly increase the manufacturing cost as well as may hamper various safety features essential for sound and uninterrupted operation of a nuclear power reactor. On the other hand, traditional heat transfer fluids such as water, ethylene glycol and oils have inherently low thermal conductivity relative to metals and even metal oxides. In this study the coolant with suspended nano-sized particles in the base fluid is proposed as an alternative to increase heat transfer but minimize flow resistance inside a nuclear reactor core. Due to technical complexities most of the previous studies carried out on heat transfer of suspension of metal oxides in fluids were limited to suspensions with millimeter or micron-sized particles. Such outsized particles may lead to severe problems in heat transfer equipment including increased pressure drop and corrosion and erosion of components and pipe lines. Dramatic advancement in modern science has made it possible to produce ultrafine metallic or nonmetallic particles of nanometer dimension, which has brought a revolutionary change in the research of heat transfer enhancement methods. Due to very tiny particle size and their small volume fraction, problems such as clogging and increased pressure drop are insignificant for nanofluids. Moreover, the relatively large surface area of nanoparticles augments the stability of nanofluid solution and prevents the sedimentation of nanoparticles. Xuan and Roetzel considered two approaches to illustrate

  5. Development of design technology on thermal-hydraulic performance in tight-lattice rod bundle. 4. Large paralleled simulation by the advanced two-fluid model code

    International Nuclear Information System (INIS)

    Misawa, Takeharu; Yoshida, Hiroyuki; Akimoto, Hajime

    2008-01-01

    In Japan Atomic Energy Agency (JAEA), the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been developed. For thermal design of FLWR, it is necessary to develop analytical method to predict boiling transition of FLWR. Japan Atomic Energy Agency (JAEA) has been developing three-dimensional two-fluid model analysis code ACE-3D, which adopts boundary fitted coordinate system to simulate complex shape channel flow. In this paper, as a part of development of ACE-3D to apply to rod bundle analysis, introduction of parallelization to ACE-3D and assessments of ACE-3D are shown. In analysis of large-scale domain such as a rod bundle, even two-fluid model requires large number of computational cost, which exceeds upper limit of memory amount of 1 CPU. Therefore, parallelization was introduced to ACE-3D to divide data amount for analysis of large-scale domain among large number of CPUs, and it is confirmed that analysis of large-scale domain such as a rod bundle can be performed by parallel computation with keeping parallel computation performance even using large number of CPUs. ACE-3D adopts two-phase flow models, some of which are dependent upon channel geometry. Therefore, analyses in the domains, which simulate individual subchannel and 37 rod bundle, are performed, and compared with experiments. It is confirmed that the results obtained by both analyses using ACE-3D show agreement with past experimental result qualitatively. (author)

  6. Steady state transient analysis of spent nuclear fuel bundle exposed to stagnant gaseous atmosphere (Paper No. HMT-56-87)

    International Nuclear Information System (INIS)

    Pal, G.; Markandeya, S.G.; Venkatraj, V.

    1987-01-01

    This paper deals with the development of a computer code for the analysis of radiative heat exchange in rod bundles. Nuclear fuel bundles continue to generate heat even after their removal from the reactor core because of decay of fission products. During the transfer of the bundles from the core to storage bay they may pass through gaseous environment. Radiative heat exchange will be the dominant mode within the bundle under this condition. A computer code RIIEINA (Radiative Heat Exchange In Nuclear Assemblies) has been developed and used for predicting the behaviour of the spent fuel subassembly of the proposed Prototype Fast Breeder Reactor exposed to gaseous environment. The analytical model computer code and the results obtained are briefly discussed. (author). 5 refs., 5 figs

  7. Numerical study of turbulent heat transfer along a heated rod in an annular cavity

    International Nuclear Information System (INIS)

    Batta, A.; Class, A.; Daubner, M.; Gnieser, S.; Stieglitz, R.

    2008-01-01

    Fundamental knowledge on the turbulent convective heat transfer from a rod into liquid metal is of crucial importance for the design of advanced liquid metal operated nuclear systems since a single rod is the basic element of a fuel rod assembly. Therefore, a numerical investigation of the heated rod experiment at KALLA (KArlsruhe Liquid metal LAboratory) has been performed. This experiment investigates the turbulent heat transfer from a heated rod placed concentrically within in a cylindrical tube in a developing flow of a heavy liquid metal (HLM, here Pb 45 Bi 55 Eutectic) at reactor typical power levels and dimensions. It is set up with thermocouples (TCs), a traversable Pitot tube and three thermocouple rakes consisting of numerous thermocouples (TCs). The concentricity is ensured by means of mechanical spacers placed axially equidistant. This article concentrates on the numerical investigation of the impact of the experimental instrumentation on the developing flow pattern and temperature field. In particular, the influence of spacers which distort the velocity profile as well of a potential contact of the spacer with the heated rod changing the heat conduction regime are considered numerically in this paper using the STAR-CD code. The turbulent flow simulation assumes axis-symmetry and uses the SST turbulence model. The simulation results exhibit a flow pattern that is substantially altered by spacers. Hence, the flow can not be considered to be axis-symmetric. This in turn yields that the convective heat transfer from the heated rod towards the spacer region is reduced leading to a temperature rise in spacer region, which represents the maximum value in this domain. As a consequence the entire three-dimensional test section must be modelled in order to correctly represent the physics and to allow an adequate interpretation of the experimental data. (orig.)

  8. Development of Design Technology on Thermal-Hydraulic Performance in Tight-Lattice Rod Bundles: I-Master Plan and Executive Summary

    Science.gov (United States)

    Ohnuki, Akira; Kureta, Masatoshi; Yoshida, Hiroyuki; Tamai, Hidesada; Liu, Wei; Misawa, Takeharu; Takase, Kazuyuki; Akimoto, Hajime

    R&D project to investigate thermal-hydraulic performance in tight-lattice rod bundles for Innovative Water Reactor for Flexible Fuel Cycle has been progressed at Japan Atomic Energy Agency in collaboration with power utilities, reactor vendors and universities since 2002. In this series-study, we will summarize the R&D achievements using large-scale test facility (37-rod bundle with full-height and full-pressure), model experiments and advanced numerical simulation technology. This first paper described the master plan for the development of design technology and showed an executive summary for this project up to FY2005. The thermal-hydraulic characteristics in the tight-lattice configuration were investigated and the feasibility was confirmed based on the experiments. We have developed the design technology including 3-D numerical simulation one to evaluate the effects of geometry/scale on the thermal-hydraulic behaviors.

  9. Thermalhydraulic behavior of electrically heated rod during a critical heat flux transient

    International Nuclear Information System (INIS)

    Lima, Rita de Cassia Fernandes de; Carajilescov, Pedro

    1997-01-01

    In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed using electrically heated rods to simulate operational and accidental conditions of nuclear fuel rods. In the present work, a theoretical analysis of the drying and rewetting front propagation is performed during a critical heat flux experiment, starting with the application of slope of electrical power from steady state condition. After the occurrence of critical heat flux, the drying front propagation is predicted. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. Studies done with several values of coolant mass flow rate show that this variable has more influence on the drying front velocity than on the rewetting one. (author)

  10. Critical heat flux experiments in tight lattice core

    Energy Technology Data Exchange (ETDEWEB)

    Kureta, Masatoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    Fuel rods of the Reduced-Moderation Water Reactor (RMWR) are so designed to be in tight lattices as to reduce moderation and achieve higher conversion ratio. As for the BWR type reactor coolant flow rate is reduced small compared with the existing BWR, so average void fraction comes to be langer. In order to evaluate thermo hydraulic characteristics of designed cores, critical heat flux experiments in tight lattice core have been conducted using simulated high pressure coolant loops for both the PWR and BWR seven fuel rod bundles. Experimental data on critical heat flux for full bundles have been accumulated and applied to assess the critical power of designed cores using existing codes. Evaluated results are conservative enough to satisfy the limiting condition. Further experiments on axial power distribution effects and 37 fuel rod bundle tests will be performed to validate thermohydraulic characteristics of designed cores. (T. Tanaka)

  11. Critical heat flux experiments in tight lattice core

    International Nuclear Information System (INIS)

    Kureta, Masatoshi

    2002-01-01

    Fuel rods of the Reduced-Moderation Water Reactor (RMWR) are so designed to be in tight lattices as to reduce moderation and achieve higher conversion ratio. As for the BWR type reactor coolant flow rate is reduced small compared with the existing BWR, so average void fraction comes to be langer. In order to evaluate thermo hydraulic characteristics of designed cores, critical heat flux experiments in tight lattice core have been conducted using simulated high pressure coolant loops for both the PWR and BWR seven fuel rod bundles. Experimental data on critical heat flux for full bundles have been accumulated and applied to assess the critical power of designed cores using existing codes. Evaluated results are conservative enough to satisfy the limiting condition. Further experiments on axial power distribution effects and 37 fuel rod bundle tests will be performed to validate thermohydraulic characteristics of designed cores. (T. Tanaka)

  12. Experimental study of heat transfer in a transverse flow around the heat exchanger tubes bank by lead

    International Nuclear Information System (INIS)

    Berezin, A.N.; Grabezhnaya, V.A.; Mikheev, A.S.; Parfenov, A.S.

    2014-01-01

    The results of the work to determine the heat transfer coefficient in crossflow by lead of pipes are presented. The study was conducted at supercritical pressure in the water circuit. There was a significant inequality in the distribution of the heat flow in different rows of the bundle of heat exchange tubes of corridor location at crossflow their lead. The experimentally determined heat transfer coefficients from the lead differ substantially from those generally accepted recommendations for the calculation of heat transfer at cross flow of rod bundle by liquid metal. The experimental results are close to those obtained earlier on the model with cross flow of heat exchanger tubes bundle by lead alloy with bismuth [ru

  13. Subchannel analysis of 37-rod tight-lattice bundle experiments for reduced-moderation water reactor

    International Nuclear Information System (INIS)

    Nakatsuka, Toru; Tamai, Hidesada; Akimoto, Hajime

    2005-01-01

    R and D project to investigate thermal-hydraulic performance of tight-lattice fuel bundles for Reduced-Moderation Water Reactor (RMWR) started at Japan Atomic Energy Research Institute (JAERI) in collaboration with utilities, reactor vendors and universities from 2002. The RMWR realizes a high conversion ratio larger than 0.1 for sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The reactor core comprises tight-lattice fuel assemblies with gap clearance of around 1.0 mm to reduce the water volume ratio to achieve the high conversion ratio. A problem of utmost importance from a thermal-hydraulic point of view is the coolability of the tight-lattice assembly with such a small gap width. JAERI has been carrying out experimental study to investigate the system parameter effects on the thermal-hydraulic performance and to confirm the feasibility of the core. In the present study, the subchannel analysis code NASCA was applied to 37-rod tight-lattice bundle experiments. The NASCA can give good predictions of critical power for the gap width of 1.3 mm while the prediction accuracy decreases for the gap width of 1.0 mm. To improve the prediction accuracy, the code will be modified to take the effect of film thickness distribution around fuel rods on boiling transition. (author)

  14. Turbulent interchange in triangular array bare rod bundles

    International Nuclear Information System (INIS)

    Kelly, J.M.; Todreas, N.

    1977-07-01

    Bulk mixing coefficients were measured for single plane water flow in a simulated rod bundle with a pitch to diameter ratio of 1.10. A tracer technique employing Rhodamine B as the tracer and measuring fluorescence was used. Isokinetic sampling was achieved by using a pressure balance method. The results were corrected for both entrance effects and diversion crossflows. The results showed a change in Reynolds number behavior as the laminar sublayer began to ''choke'' the turbulent mixing. This, and a review of other mixing experiments, suggested that secondary flows do not compensate for laminarization and that turbulent mixing decreases as the pitch to diameter ratio decreases for values of P/D less than 1.05 in a manner similar to that predicted by Ramm et al. Concentration profiles were measured through the clearance gap and the values of the gradient were used to calculate the gap averaged circumferential eddy diffusivity for mass. A discussion of the eddy diffusivity concept and its applicability to turbulent mixing is presented

  15. On turbulence models for rod bundle flow computations

    International Nuclear Information System (INIS)

    Hazi, Gabor

    2005-01-01

    In commercial computational fluid dynamics codes there is more than one turbulence model built in. It is the user responsibility to choose one of those models, suitable for the problem studied. In the last decade, several computations were presented using computational fluid dynamics for the simulation of various problems of the nuclear industry. A common feature in a number of those simulations is that they were performed using the standard k-ε turbulence model without justifying the choice of the model. The simulation results were rarely satisfactory. In this paper, we shall consider the flow in a fuel rod bundle as a case study and discuss why the application of the standard k-ε model fails to give reasonable results in this situation. We also show that a turbulence model based on the Reynolds stress transport equations can provide qualitatively correct results. Generally, our aim is pedagogical, we would like to call the readers attention to the fact that turbulence models have to be selected based on theoretical considerations and/or adequate information obtained from measurements

  16. Subchannel Scale Thermal-Hydraulic Analysis of Rod Bundle Geometry under Single-phase Adiabatic Conditions Using CUPID

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Seok Jong; Park, Goon Cherl; Cho, Hyoung Kyu [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In Korea, subchannel analysis code, MATRA has been developed by KAERI (Korea Atomic Energy Research Institute). MATRA has been used for reactor core T/H design and DNBR (Departure from Nucleate Boiling Ratio) calculation. Also, the code has been successfully coupled with neutronics code and fuel analysis code. However, since major concern of the code is not the accident simulation, some features of the code are not optimized for the accident conditions, such as the homogeneous model for two-phase flow and spatial marching method for numerical scheme. For this reason, in the present study, application of CUPID for the subchannel scale T/H analysis in rod bundle geometry was conducted. CUPID is a component scale T/H analysis code which adopts three dimensional two-fluid three-field model developed by KAERI. In this paper, the validation results of the CUPID code for subchannel scale rod bundle analysis at single phase adiabatic conditions were presented. At first, the physical models required for a subchannel scale analysis were implemented to CUPID. In the future, the scope of validation tests will be extended to diabetic and two phase flow conditions and required models will be implemented into CUPID.

  17. Quasi-steady state boiling downstream of a central blockage in a 19-rod simulated LMFBR subassembly (FFM bundle 3B)

    International Nuclear Information System (INIS)

    Hanus, N.; Fontana, M.H.; Gnadt, P.A.; MacPherson, R.E.; Smith, C.M.; Wantland, J.L.

    1976-01-01

    Results of sodium boiling tests in a centrally blocked 19-rod simulated LMFBR subassembly are discussed. The tests were part of the experimental series conducted with bundle 3B in the Fuel Failure Mockup (FFM) at ORNL

  18. Transient void fraction measurements in rod bundle geometries

    International Nuclear Information System (INIS)

    Chan, A.M.C.

    1998-01-01

    A new gamma densitometer with a Ba-133 source and a Nal(TI) scintillator operated in the count mode has been designed for transient void fraction measurements in the RD-14M heated channels containing a seven-element heater bundle. The device was calibrated dynamically in the laboratory using an air-water flow loop. The void fraction measured was found to compare well with values obtained using the trapped-water method. The device was also found to follow very well the passage of air slugs in pulsating flow with slug passing frequencies of up to about 1.5 hz. (author)

  19. SIMPLE-2: a computer code for calculation of steady-state thermal behavior of rod bundles with flow sweeping

    International Nuclear Information System (INIS)

    Jones, O.C. Jr.; Yao, S.; Henry, R.E.

    1976-01-01

    A computer code has been developed for use in making single-phase thermal hydraulic calculations in rod bundle arrays with flow sweeping due to spiral wraps as the predominant crossflow mixing effect. This code, called SIMPLE-2, makes the assumption that the axial pressure gradient is identical for each subchannel over a given axial increment, and is unique in that no empirical coefficients must be specified for its use. Results from this code have been favorably compared with experimental data for both uniform and highly nonuniform power distributions. Typical calculations for various bundle sizes applicable to the LMBR program are included

  20. Critical Heat Flux Phenomena at HighPressure & Low Mass Fluxes: NEUP Final Report Part I: Experiments

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-30

    This report is a preliminary document presenting an overview of the Critical Heat Flux (CHF) phenomenon, the High Pressure Critical Heat Flux facility (HPCHF), preliminary CHF data acquired, and the future direction of the research. The HPCHF facility has been designed and built to study CHF at high pressure and low mass flux ranges in a rod bundle prototypical of conceptual Small Modular Reactor (SMR) designs. The rod bundle is comprised of four electrically heated rods in a 2x2 square rod bundle with a prototypic chopped-cosine axial power profile and equipped with thermocouples at various axial and circumferential positions embedded in each rod for CHF detection. Experimental test parameters for CHF detection range from pressures of ~80 – 160 bar, mass fluxes of ~400 – 1500 kg/m2s, and inlet water subcooling from ~30 – 70°C. The preliminary data base established will be further extended in the future along with comparisons to existing CHF correlations, models, etc. whose application ranges may be applicable to the conditions of SMRs.

  1. Enhancement of nuclear heat transfer in a typical pressurized water reactor by new spacer grids

    International Nuclear Information System (INIS)

    Nazifi, M.; Nematollahi, M.

    2007-01-01

    The fuel element geometry typically used in nuclear reactor is rod bundle whose rod-to-rod clearance is maintained by grid spacer. The heat generated in the rod by nuclear reaction is removed by coolant, usually in turbulent flow. The coolant moves axially through the subchannels. Fuel spacer grid affects the coolant flow distribution in a fuel rod bundle, and so spacer geometry has a strong influence on a bundle's thermal-hydraulic characteristics such as critical heat flux and pressure drop. An understanding of the detailed structure of the turbulent flow and heat transfer in the rod bundle, used especially as nuclear fuel elements, is of major interest to the nuclear power industry for their safe and reliable operation. The flow mixing devices on grid spacer would enhance the mixing rate between sub-channels and promote the turbulence in subchannel. The present study evaluates the effects of mixing vane shape on flow structure and heat transfer downstream of mixing vane in a sub-channel of fuel assembly, by obtaining velocity and pressure fields, turbulent intensity, flow mixing factors, heat transfer coefficient and friction factor using three-dimensional RANS analysis. Six new shapes mixing vane designed by the authors, are simulated numerically to evaluate the performance in enhancing the heat transfer, in comparison with commercialized split vane. Standard K-epsilon model are used as a turbulence closure model and periodic and symmetry condition are set as boundary conditions. The capability of the model to predict the coolant flow distribution inside rod bundles is shown and discussed on the base of comparison with experimental data for a variety of geometrical and Reynolds number conditions. It is conformed that the turbulence in the sub-channel was significantly promoted by spacer and mixing devices but rapidly decreased to a fully developed level approximately 10 time of hydraulic diameter downstream of the top of spacer. Ring type mixer showed a high

  2. Energy-1: a computer code for thermohydraulic analysis of a LMBFR rod bundles, in a mixed convection regime

    International Nuclear Information System (INIS)

    Braz Filho, F.A.

    1987-01-01

    A code was set up in which velocity, temperature and pressure distributions are calculated, using the porous body model, for a rod bundle where mixed convection regime plays an important role. Results show satisfactory agreement with experimental data, as well as a reduction in computational time when compared to ENERGY-III code. (author) [pt

  3. An electron-beam-heating model for the Gamble II rod pinch

    International Nuclear Information System (INIS)

    Mosher, David; Schumer, Joseph; Hinshelwood, David; Weber, Bruce; Stephanakis, Stavros; Swanekamp, Stephen; Young, Frank

    2002-01-01

    The rod-pinch diode concentrates electron deposition onto the tip of a high-atomic-number, mm-dia. anode rod to create an ultra-bright x-ray source for multi-MV radiography. Here, a technique is presented whereby line-spread functions acquired on-axis and at 90 deg. to the rod are used to determine the electron-deposition distribution. Results show that the smaller measured on-axis spot size for heated rods on Gamble II is due to pinching closer to the tapered tip. For a diode power of 6x1010 W, peak electron heating of 1x1014 W/cm 3 is calculated. MHD calculations of the e-beam-heated rod response agree with Schlieren measurements of plasma expansion

  4. Boiling on a tube bundle: heat transfer, pressure drop and flow patterns

    International Nuclear Information System (INIS)

    Royen Van, E.

    2011-11-01

    The complexity of two-phase flow boiling on a tube bundle presents many challenges to the understanding of the physical phenomena taking place. It is important to quantify these numerous heat flow mechanisms in order to better describe the performance of tube bundles as a function of the operational conditions. In the present study, the bundle boiling facility at the Laboratory of Heat and Mass Transfer (LTCM) was modified to obtain high-speed videos to characterise the two-phase regimes and some bubble dynamics of the boiling process. It was then used to measure heat transfer on single tubes and in bundle boiling conditions. Pressure drop measurements were also made during adiabatic and diabatic bundle conditions. New enhanced boiling tubes from Wolverine Tube Inc. (Turbo-B5) and the Wieland-Werke AG (Gewa-B5) were investigated using R134a and R236fa as test fluids. The tests were carried out at saturation temperatures T sat of 5 °C and 15 °C, mass flow rates from 4 to 35 kg/m 2 s and heat fluxes from 15 to 70 kW/m 2 , typical of actual operating conditions. The flow pattern investigation was conducted using visual observations from a borescope inserted in the middle of the bundle. Measurements of the light attenuation of a laser beam through the intertube two-phase flow and local pressure fluctuations with piezo-electric pressure transducers were also taken to further help in characterising the complex flow. Pressure drop measurements and data reduction procedures were revised and used to develop new, improved frictional pressure drop prediction methods for adiabatic and diabatic two-phase conditions. The physical phenomena governing the enhanced tube evaporation process and their effects on the performance of tube bundles were investigated and insight gained. A new method based on a theoretical analysis of thin film evaporation was used to propose a new correlating parameter. A large new database of local heat transfer coefficients were obtained and then

  5. Experimental study on the Reynolds number dependence of turbulent mixing in a rod bundle

    International Nuclear Information System (INIS)

    Silin, Nicolas; Juanico, Luis

    2006-01-01

    An experimental study for Reynolds number dependence of the turbulent mixing between fuel-bundle subchannels, was performed. The measurements were done on a triangular array bundle with a 1.20 pitch to diameter relation and 10 mm rod diameter, in a low-pressure water loop, at Reynolds numbers between 1.4 x 10 3 and 1.3 x 10 5 . The high accuracy of the results was obtained by improving a thermal tracing technique recently developed. The Reynolds exponent on the mixing rate correlation was obtained with two-digit accuracy for Reynolds numbers greater than 3 x 10 3 . It was also found a marked increase in the mixing rate for lower Reynolds numbers. The weak theoretical base of the accepted Reynolds dependence was pointed out in light of the later findings, as well as its ambiguous supporting experimental data. The present results also provide indirect information about dominant large scale flow pulsations at different flow regimes

  6. Multidimensional simulations of fuel rod appendage effects on pressure drop and heat transfer in an annulus flow

    International Nuclear Information System (INIS)

    Banas, A.O.; Carver, M.B.; Leung, J.C.H.; Bromley, B.P.

    1992-10-01

    The general purpose computational fluid dynamics code, Harwell-FLOW3D, has been used to simulate the effects of fuel rod obstructions on pressure drop and heat transfer in single phase turbulent flows in a concentric annular channel. The results of two and three dimensional simulations are reported for obstructions approximating the geometry of bearing pads used in 37 element CANDU fuel bundles. Pressure drop penalty and augmentation of heat transfer have been quantified and correlated with the obstruction geometrical parameters and the dimensionless numbers representing operating conditions. The predicted effects on pressure drop have been compared with several experimental correlations, yielding good agreement. The methodology presented offers results that can be used directly as input into thermalhydraulic analyses in subchannel and system codes. (Author) (23 figs., 15 refs.)

  7. Thermalhydraulic behavior of electrically heated rods during critical heat flux transients

    International Nuclear Information System (INIS)

    Lima, Rita de Cassia Fernandes de

    1997-01-01

    In nuclear reactors, the occurrence of critical heat flux leads to fuel rod overheating with clad fusion and radioactive products leakage. To predict the effects of such phenomenon, experiments are performed utilizing heated rods to simulate operational and accidental conditions of nuclear fuel rods, with special attention to the phenomenon of boiling crisis. The use of mechanisms which detect the abrupt temperature rise allows the electric power switch off. These facts prevent the test section from damage. During the critical heat flux phenomenon the axial heat conduction becomes very important. The study of the dryout and rewetting fronts yields the analysis, planning and following of critical heat flux experiments. These facts are important during the reflooding of nuclear cores at severe accidents. In the present work it is performed a theoretical analysis of the drying and rewetting front propagation during a critical heat flux experiment, starting with the application of an electrical power step or power slope from steady state condition. After the occurrence of critical heat flux, it is predicted the drying front propagation. After a few seconds, a power cut is considered and the rewetting front behavior is analytically observed. In all these transients the coolant pressure is 13,5 MPa. For one of them, comparisons are done with a pressure of 8,00 MPa. Mass flow and enthalpy influences on the fronts velocities are also analysed. These results show that mass flow has more importance on the drying front velocities whereas the pressure alters strongly the rewetting ones. (author)

  8. Steady-state and transient studies on critical heat flux of a PWR 5 x 5 fuel element bundle with complex spacer wire geometry

    International Nuclear Information System (INIS)

    Fulfs, H.; Katsaounis, A.; Kreubig, M.; Minden, C. von; Orlowski, R.

    1980-01-01

    The results will be described in exemplary presentations completely and concluding. The experimental examination of the steady state simularity of critical heat flux (CHF) in freon 12 and water at identical PWR-5 x 15-rod bundles will show that hot rod/hot channels position as well as CHF can be transformed from model to original fluid with good accuracy. The investigated mass flow and power transients (only in freon 12) point out a definite influence of initial and boundary conditions on CHF and CHF time delay at changing rates higher than 10 to 20%/s. On the contrary simulation of primary pump failure (LOFA) shows no or only small improvement in CHF behaviour while a coupled Scram prevents from reaching the boiling crisis. (orig.) [de

  9. 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

  10. Wire-wrap bundle compression-characteristics study. Phase I

    International Nuclear Information System (INIS)

    Chertock, A.J.

    1974-06-01

    An analytical computer comparison was made of the compression characteristics of proposed wire-wrap bundles. The study included analysis of 7- and 37-rod straight-start bundles (base configuration), and softened 37-rod configurations. The softened configurations analyzed were: straight-start with distributed wireless fuel rods, and the staggered wire-wrap start angles of 0 0 -30 0 -60 0 and 0 0 -45 0 -90 0 . The compression of the bundle simulates the bundle-to-channel interference at end-of-life conditions at which high differential swelling between the channel and bundle has been predicted. The computer results do not include the so-called dispersion effects. The effects of other variables such as pitch length, creep, axial variations in swelling, and degree of swelling were not studied. These analytic studies give an indication of trends only. No credence should be given to specific quantitative load or deflection results quoted in this report

  11. Experimental determination of local heat flux variation in an electrically heated BR-2 rod

    International Nuclear Information System (INIS)

    Meyer, L.; Merschroth, F.

    1977-08-01

    The installation of thermocouples within the cladding of an electrically heated BR-2 rod might cause local variations of heat flux. In order to detect a resulting temperature variation at the outer surface, experiments with a single electrically heated rod with heat fluxes up to 30.80 W/cm 2 and heat transfer coefficients up to 1000 W/m 2 K by forced convection in air were conducted. The surface temperatures were measured with an optical pyrometer. The experiment showed about 0.6% variation in the surface temperature. An analysis with the TAC2D-code shows that local variation in the heat flux under these conditions is less than 1.2%. (orig.) [de

  12. 5 X 5 rod bundle flow field measurements downstream a PWR spacer grid

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Higor F.P.; Silva, Vitor V A.; Santos, André A.C.; Veloso, Maria A.F., E-mail: higorfabiano@gmail.com, E-mail: mdora@nuclear.ufmg.br, E-mail: vitors@cdtn.br, E-mail: aacs@cdtn.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil); Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The spacer grids are structures present in nuclear fuel assembly of Pressurized Water Reactors (PWR). They play an important structural role and also assist in heat removal through the assembly by promoting increased turbulence of the flow. Understanding the flow dynamics downstream the spacer grid is paramount for fuel element design and analysis. This paper presents water flow velocity profiles measurements downstream a spacer grid in a 5 x 5 rod bundle test rig with the objective of highlighting important fluid dynamic behavior near the grid and supplying data for CFD simulation validation. These velocity profiles were obtained at two different heights downstream the spacer grid using a LDV (Laser Doppler Velocimetry) through the top of test rig. The turbulence intensities and patterns of the swirl and cross flow were evaluated. The tests were conducted for Reynolds numbers ranging from 1.8 x 10{sup 4} to 5.4 x 10{sup 4}. This experimental research was carried out in thermo-hydraulics laboratory of Nuclear Technology Development Center – CDTN. Results show great repeatability and low uncertainties (< 1.24 %). Details of the flow field show how the mixture and turbulence induced by the spacer grid quickly decays downstream the spacer grid. It is shown that the developed methodology can supply high resolution low uncertainty results that can be used for validation of CFD simulations. (author)

  13. 5 X 5 rod bundle flow field measurements downstream a PWR spacer grid

    International Nuclear Information System (INIS)

    Castro, Higor F.P.; Silva, Vitor V A.; Santos, André A.C.; Veloso, Maria A.F.

    2017-01-01

    The spacer grids are structures present in nuclear fuel assembly of Pressurized Water Reactors (PWR). They play an important structural role and also assist in heat removal through the assembly by promoting increased turbulence of the flow. Understanding the flow dynamics downstream the spacer grid is paramount for fuel element design and analysis. This paper presents water flow velocity profiles measurements downstream a spacer grid in a 5 x 5 rod bundle test rig with the objective of highlighting important fluid dynamic behavior near the grid and supplying data for CFD simulation validation. These velocity profiles were obtained at two different heights downstream the spacer grid using a LDV (Laser Doppler Velocimetry) through the top of test rig. The turbulence intensities and patterns of the swirl and cross flow were evaluated. The tests were conducted for Reynolds numbers ranging from 1.8 x 10"4 to 5.4 x 10"4. This experimental research was carried out in thermo-hydraulics laboratory of Nuclear Technology Development Center – CDTN. Results show great repeatability and low uncertainties (< 1.24 %). Details of the flow field show how the mixture and turbulence induced by the spacer grid quickly decays downstream the spacer grid. It is shown that the developed methodology can supply high resolution low uncertainty results that can be used for validation of CFD simulations. (author)

  14. Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 61-Rod Bundle for a Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Hyungmo; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Jeong, Ji-Young; Lee, Hyeong-Yeon

    2015-01-01

    For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are crucial factor for the design code verification and validation. Wrapped wires make a cross flow in a circumference of the fuel rod, and this effect lets flow be mixed. Therefore the sub-channel analysis method is commonly used for thermal hydraulic analysis of a SFR, a wire wrapped sub-channel type. To measure flow mixing characteristics, a wire mesh sensing technique can be useful method. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, the recent reports that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. The subchannel flow characteristics analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped subchannel type. In this study, mixing experiments were conducted successfully at a hexagonally arrayed 61-pin wire-wrapped fuel rod bundle test section. Wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable

  15. Boiling on a tube bundle: heat transfer, pressure drop and flow patterns

    International Nuclear Information System (INIS)

    Agostini, F.

    2008-07-01

    The complexity of the two-phase flow in a tube bundle presents important problems in the design and understanding of the physical phenomena taking place. The working conditions of an evaporator depend largely on the dynamics of the two-phase flow that in turn influence the heat exchange and the pressure drop of the system. A characterization of the flow dynamics, and possibly the identification of the flow pattern in the tube bundle, is thus expected to lead to a better understanding of the phenomena and to reveal on the mechanisms governing the tube bundle. Therefore, the present study aims at providing further insights into two-phase bundle flow through a new visualization system able to provide for the first time a view of the flow in the core of a tube bundle. In addition, the measurement of the light attenuation of a laser beam through the two-phase flow and measurement of the high frequency pressure fluctuations with a piezo-electric pressure transducer are used to characterize the flow. The design and the validation of this new instrumentation also provided a method for the detection of dry-out in tube bundles. This was achieved by a laser attenuation technique, flow visualization, and estimation of the power spectrum of the pressure fluctuation. The current investigation includes results for two different refrigerants, R134a and R236fa, three saturations temperatures T sat = 5, 10 and 15 °C, mass velocities ranging from 4 to 40 kg/sm² in adiabatic and diabatic conditions (several heat fluxes). Measurement of the local heat transfer coefficient and two-phase frictional pressure drop were obtained and utilized to improve the current prediction methods. The heat transfer and pressure drop data were supported by extensive characterization of the two-phase flow, which was to improve the understanding of the two-phase flow occurring in tube bundles. (author)

  16. Measurements of peripherical static pressure and pressure drop in a rod bundle with helical wire wrap spacers

    International Nuclear Information System (INIS)

    Ballve, H.; Graca, M.C.; Fernandez y Fernandez, E.; Carajilescov, P.

    1981-07-01

    The fuel element of a LMFBR nuclear reactor consists of a wire wrapped rod bundle with triangular array with the coolant flowing parallel to the rods. Using this type of element with seven rods conected to an air open loop. The hydrodinamics behavior of the flow for p/d = 1.20 and l/d = 15.0, was simulated. Several measurements were performed in order to obtain the static pressure distribution at the walls of the hexagonal duct, for Reynolds number from 4.4x10 3 to 48.49x10 3 and for different axial and transverse positions, in a wire wrap lead. The axial pressure drop was obtained and determined the friction factor dependence with the Reynolds number. From the obtained results, it was observed the non-dependency of the non-dimensionalized axial and transverse local static pressure distribution at the wall of the hexagonal duct, with the Reynolds number. The obtained friction factor is compared to the results of previous works. (Author) [pt

  17. Severe fuel damage experiments performed in the QUENCH facility with 21-rod bundles of LWR-type

    International Nuclear Information System (INIS)

    Sepold, L.; Hering, W.; Schanz, G.; Scholtyssek, W.; Steinbrueck, M.; Stuckert, J.

    2006-01-01

    The objective of the QUENCH experimental program at the Karlsruhe Research Center is to investigate core degradation and the hydrogen source term that results from quenching/flooding an uncovered core, to examine the physical/chemical behavior of overheated fuel elements under different flooding conditions, and to create a data base for model development and improvement of severe fuel damage (SFD) code systems. The large-scale 21-rod bundle experiments conducted in the QUENCH out-of-pile facility are supported by an extensive separate-effects test program, by modeling activities as well as application and improvement of SFD code systems. International cooperations exist with institutions mainly within the European Union but e.g. also with the Russian Academy of Science (IBRAE, Moscow) and the CSARP program of the USNRC. So far, eleven experiments have been performed, two of them with B 4 C absorber material. Experimental parameters were: the temperature at initiation of reflood, the degree of peroxidation, the quench medium, i.e. water or steam, and its injection rate, the influence of a B 4 C absorber rod, the effect of steam-starved conditions before quench, the influence of air oxidation before quench, and boil-off behavior of a water-filled bundle with subsequent quenching. The paper gives an overview of the QUENCH program with its organizational structure, describes the test facility and the test matrix with selected experimental results. (author)

  18. Performance analysis of LMFBR control rods

    International Nuclear Information System (INIS)

    Pitner, A.L.; Birney, K.R.

    1975-01-01

    Control rods in the FFTF and LMFBR's will consist of pin bundles of stainless steel-clad boron carbide pellets. In the FFTF reference design, sixty-one pins of 0.474-inch diameter each containing a 36-inch stack of 0.362-inch diameter boron carbide pellets comprise a control rod. Reactivity control is provided by the 10 B (n,α) 7 Li reaction in the boron carbide. This reaction is accompanied by an energy release of 2.8 MeV, and heating from this reaction typically approaches 100 watts/cm 3 for natural boron carbide pellets in an LMFBR flux. Performance analysis of LMFBR control rods must include an assessment of the thermal performance of control pins. In addition, irradiation performance with regard to helium release, pellet swelling, and reactivity worth depletion as a function of service time must be evaluated

  19. Acoustic loading effects on oscillating rod bundles

    International Nuclear Information System (INIS)

    Lin, W.H.

    1980-01-01

    An analytical study of the interaction between an infinite acoustic medium and a cluster of circular rods is described. The acoustic field due to oscillating rods and the acoustic loading on the rods are first solved in a closed form. The acoustic loading is then used as a forcing function for rod responses, and the acousto-elastic couplings are solved simultaneously. Numerical examples are presented for several cases to illustrate the effects of various system parameters on the acoustic reaction force coefficients. The effect of the acoustic loading on the coupled eigenfrequencies are discussed

  20. Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 37-Rod Bundle for a Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Hyungmo; Bae, Hwang; Chang, Seok-Kyu; Choi, Sun Rock; Lee, Dong Won; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Lee, Hyeong-Yeon

    2014-01-01

    For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are very important. Wrapped wires make a cross flow in a around the fuel rod) of the fuel rod, and this effect lets flow be mixed. Experimental results of flow mixing can be meaningful for verification and validation of thermal mixing correlation in a reactor core thermo-hydraulic design code. A wire mesh sensing technique can be useful method for measuring of flow mixing characteristics. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, it has been recently reported that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. This can be powerfully adapted to recognize flow mixing characteristics by wrapped wires in SFR core thermal design. In this work, we conducted the flow mixing experiments using a custom designed wire mesh sensor. To verify and validate computer codes for the SFR core thermal design, mixing experiments were conducted at a hexagonally arrayed 37-pin wire-wrapped fuel rod bundle test section. The well-designed wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable. In addition, by uncertainty analysis, the system errors and the random error were estimated in experiments. Therefore, the present results and methods can be used for design code verification and validation

  1. A Validation of Subchannel Based CHF Prediction Model for Rod Bundles

    International Nuclear Information System (INIS)

    Hwang, Dae-Hyun; Kim, Seong-Jin

    2015-01-01

    A large number of CHF data base were procured from various sources which included square and non-square lattice test bundles. CHF prediction accuracy was evaluated for various models including CHF lookup table method, empirical correlations, and phenomenological DNB models. The parametric effect of the mass velocity and unheated wall has been investigated from the experimental result, and incorporated into the development of local parameter CHF correlation applicable to APWR conditions. According to the CHF design criterion, the CHF should not occur at the hottest rod in the reactor core during normal operation and anticipated operational occurrences with at least a 95% probability at a 95% confidence level. This is accomplished by assuring that the minimum DNBR (Departure from Nucleate Boiling Ratio) in the reactor core is greater than the limit DNBR which accounts for the accuracy of CHF prediction model. The limit DNBR can be determined from the inverse of the lower tolerance limit of M/P that is evaluated from the measured-to-predicted CHF ratios for the relevant CHF data base. It is important to evaluate an adequacy of the CHF prediction model for application to the actual reactor core conditions. Validation of CHF prediction model provides the degree of accuracy inferred from the comparison of solution and data. To achieve a required accuracy for the CHF prediction model, it may be necessary to calibrate the model parameters by employing the validation results. If the accuracy of the model is acceptable, then it is applied to the real complex system with the inferred accuracy of the model. In a conventional approach, the accuracy of CHF prediction model was evaluated from the M/P statistics for relevant CHF data base, which was evaluated by comparing the nominal values of the predicted and measured CHFs. The experimental uncertainty for the CHF data was not considered in this approach to determine the limit DNBR. When a subchannel based CHF prediction model

  2. Critical heat flux detection in rods simulating fuel elements by using dilation method

    International Nuclear Information System (INIS)

    Mesquita, A.Z.

    1993-01-01

    In out-reactor heat transfer experiments, fuel elements are often simulated by electrically heated rods. In order to prevent the heating rod from being damaged by burnout, when the critical heat flux occurs a safety system is provided which checks the axial thermal expansion of the rod. In case of sudden temperature increase, the corresponding elongation causes a fast interruption of the electrical power supply. The experiments presented here show that this method is more effective than one that uses thermocouples. (author)

  3. Tube spacer grid for a heat-exchanger tube bundle

    International Nuclear Information System (INIS)

    Scheidl, H.

    1976-01-01

    A tube spacer grid for a heat-exchanger tube bundle is formed by an annular grid frame having a groove formed in its inner surface in which the interspaced grid bars have their ends positioned and held in interspaced relationship by short sections of tubes passed through holes axially formed in the grid frame so that the tubes are positioned between the ends of the grid bars in the grooves. The tube sections may be cut from the same tubes used to form the tube bundle. 5 claims, 3 drawing figures

  4. Development of boiling transition analysis code TCAPE-INS/B based on mechanistic methods for BWR fuel bundles. Models and validations with boiling transition experimental data

    International Nuclear Information System (INIS)

    Ishida, Naoyuki; Utsuno, Hideaki; Kasahara, Fumio

    2003-01-01

    The Boiling Transition (BT) analysis code TCAPE-INS/B based on the mechanistic methods coupled with subchannel analysis has been developed for the evaluation of the integrity of Boiling Water Reactor (BWR) fuel rod bundles under abnormal operations. Objective of the development is the evaluation of the BT without using empirical BT and rewetting correlations needed for different bundle designs in the current analysis methods. TCAPE-INS/B consisted mainly of the drift-flux model, the film flow model, the cross-flow model, the thermal conductivity model and the heat transfer correlations. These models were validated systematically with the experimental data. The accuracy of the prediction for the steady-state Critical Heat Flux (CHF) and the transient temperature of the fuel rod surface after the occurrence of BT were evaluated on the validations. The calculations for the experiments with the single tube and bundles were carried out for the validations of the models incorporated in the code. The results showed that the steady-state CHF was predicted within about 6% average error. In the transient calculations, BT timing and temperature of the fuel rod surface gradient agreed well with experimental results, but rewetting was predicted lately. So, modeling of heat transfer phenomena during post-BT is under modification. (author)

  5. Subchannel measurements of the equilibrium quality and mass flux distribution in a rod bundle

    International Nuclear Information System (INIS)

    Lahey, R.T. Jr.

    1986-01-01

    An experiment was performed to measure the equilibrium subchannel void and mass flux distribution in a simulated BWR rod bundle. These new equilibrium subchannel data are unique and represent an excellent basis for subchannel ''void drift'' model development and assessment. Equilibrium subchannel void and mass flux distributions have been determined from the data presented herein. While the form of these correlations agree with the results of previous theoretical investigations, they should be generalized with caution since the current data base has been taken at only one (low) system pressure. Clearly there is a need for equilibrium subchannel data at higher system pressures if mechanistic subchannel models are to be developed

  6. Temperature field downstream of an heated bundle mock-up results for different power distribution

    International Nuclear Information System (INIS)

    Girard, J.P.; Buravand, Y.

    1982-10-01

    The aim of these peculiar experiments performed on the ML4 loop in ISPRA is to evaluate the characteristics of the temperature field over a length of 20 to 30 dias downstream of a rod bundle for different temperatures profiles at the bundle outlet. The final purpose of this work will be to establish either directly or through models whether it is possible or not to detect subassembly failures using suitable of the subassembly outlet temperature signal. 15 hours of digital and analog recording were taped for five different power distributions in the bundle. The total power dissipation remained constant during the whole run. Two flow rates and seven axial location were investigated. It is shown that the different temperature profiles produce slight differences in the variance and skewness of the temperature signal measured along the axis of the pipe over 20 dias

  7. Experimental data and calculation studies of critical heat fluxes at local disturbances of geometry of WWER fuel assemblies

    International Nuclear Information System (INIS)

    Kobzar, L.L.; Oleksyuk, D.A.

    2001-01-01

    The results of experiments executed in RRC 'Kurchatov Institute on the thermal-physical critical facility SVD are presented herein. The experiments modeled the drawing of two fuel rods to each other till touching WWER-1000 reactor in FA. The experimental model is a 7-rod bundle with the heated length of 1 m. The primary goal of experiments was to acquire the quantitative factors of the reduction in the critical heat fluxes as contrasted to the basic model (without disturbances of FA geometry) at the expense of local disturbance of a rod bundle geometry. As it follows from the experiment, the effect of decrease of the critical heat rate depends on combination of regime parameters and it makes 15% in the most unfavorable case (Authors)

  8. Experimental Study of Hydraulic Control Rod Drive Mechanism for Passive IN-core Cooling System of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    CAREM 25 (27 MWe safety systems using hydraulic control rod drives (CRD) studied critical issues that were rod drops with interrupted flow [3]. Hydraulic control rod drive suggested fast shutdown condition using a large gap between piston and cylinder in order to fast drop of neutron absorbing rods. A Passive IN-core Cooling system (PINCs) was suggested for safety enhancement of pressurized water reactors (PWR), small modular reactor (SMR), sodium fast reactor (SFR) in UNIST. PINCs consist of hydraulic control rod drive mechanism (Hydraulic CRDM) and hybrid control rod assembly with heat pipe combined with control rod. The schematic diagram of the hydraulic CRDM for PINCs is shown in Fig. 1. The experimental results show the steady state and transient behavior of the upper cylinder at a low pressure and low temperature. The influence of the working fluid temperature and cylinder mass are investigated. Finally, the heat removal between evaporator section and condenser section is compared with or without the hybrid control rod. Heat removal test of the hybrid heat pipe with hydraulic CRDM system showed the heat transfer coefficient of the bundle hybrid control rod and its effect on evaporator pool. The preliminary test both hydraulic CRDM and heat removal system was conducted, which showed the possibility of the in-core hydraulic drive system for application of PINCs.

  9. Vernotte-Cattaneo approximation for heat conduction in fuel rod

    International Nuclear Information System (INIS)

    Espinosa P, G.; Espinosa M, E. G.

    2009-10-01

    In this paper we explore the applicability of a fuel rod mathematical model based on the Vernotte-Cattaneo transient heat conduction as constitutive law (Non-Fourier approach) for light water reactors transient analysis. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The motivation for this research was to eliminate the paradox of an infinite. The motivation for this research was to eliminate the paradox of an infinite thermal wave speed. The time-dependent heat sources were considered in the fuel rod heat transfer model. The close of the main steam isolated valves transient in a boiling water reactor was analyzed for different relaxation times. The results show that for long-times the heat fluxes on the clad surface under Vernotte-Cattaneo approach can be important, while for short-times and from the engineering point of view the changes are very small. (Author)

  10. Natural convection heat transfer within horizontal spent nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Canaan, R.E.

    1995-12-01

    Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array

  11. Pressure drop characteristics in tight-lattice bundles for reduced-moderation water reactors

    International Nuclear Information System (INIS)

    Tamai, Hidesada; Kureta, Masatoshi; Yoshida, Hiroyuki; Akimoto, Hajime

    2004-01-01

    The reduced-moderation water reactor (RMWR) consists of several distinctive structures; a triangular tight-lattice configuration and a double-flat core. In order to design the RMWR core from the point of view of thermal-hydraulics, an evaluation method on pressure drop characteristics in the rod bundles at the tight-lattice configuration is required. In this study, calculated results by the Martinelli-Nelson's and Hancox's correlations were compared with experimental results in 4 x 5 rod bundles and seven-rod bundles. Consequently, the friction loss in two-phase flows becomes smaller at the tight-lattice configuration with the hydraulic diameter less than about 3 mm. This reason is due to the difference of the configuration between the multi-rod bundle and the circular tube and due to the effect of the small hydraulic diameter on the two-phase multiplier. (author)

  12. Stress Analysis of Fuel Rod under Axial Coolant Flow

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

    2010-05-15

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  13. Stress Analysis of Fuel Rod under Axial Coolant Flow

    International Nuclear Information System (INIS)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

    2010-01-01

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  14. Oriented heat release in asphalt pavement induced by high-thermal-conductivity rods

    International Nuclear Information System (INIS)

    Du, Yinfei; Wang, Shengyue

    2015-01-01

    In this paper, a new principle of using aligned high-thermal-conductivity rods to enhance the oriented heat conduction in asphalt pavement was proposed. The results showed that the designed structure absorbed more heat during the day. The heat flow in the designed structure presented a non-uniform horizontal distribution. At the depth of 4 cm, the horizontal and vertical heat fluxes through steel rods were thirteen and ten times higher than those through asphalt mixture, respectively. The maximum temperature of the designed structure reduced by 3.6 °C–6.5 °C at the depth of 4 cm. The results of indoor irradiation test showed a trend consistent with those of numerical simulation. After 500 thousand times of standard axis load were applied, the rutting depth of the designed structure reduced by 43.4%. The principle proposed is expected to be used to induce an oriented heat release accumulated in asphalt pavement and reduce pavement temperature and rutting. - Highlights: • Steel rods were inserted in the middle and bottom layers to build thermal channels. • Steel rods absorbed heat from asphalt mixture and rapidly released them to subgrade. • The heat flux through asphalt mixture decreased and pavement temperature reduced.

  15. Out-of-pile bundle temperature escalation under severe fuel damage conditions

    International Nuclear Information System (INIS)

    Hagen, S.; Peck, S.O.

    1983-08-01

    This report provides an overview of the test conduct, results, and posttest appearance of bundle test ESBU-1. The purpose of the test was to investigate fuel rod temperature escalation due to the exothermal zircaloy/steam reaction in a bundle geometry. The 3x3 bundle was surrounded by a zircaloy shroud and 6 mm of fiber ceramic insulation. The center rod escalated to a maximum of 2,250 0 C. Runoff of the melt apparently limited the escalation. Posttest visual examination of the bundle showed that cladding from every rod had melted, liquefied some fuel, flowed down the rod, and frozen in a solid mass that substantially blocked all flow channels. A large amount of powdery rubble, probably fuel that fractured during cooldown, was found on top of the blockage. Metallographic, EMP, and SEM examinations showed that the melt had dissolved both fuel and oxidized cladding, and had itself been oxidized by steam. (orig.) [de

  16. Triangularly arranged heat exchanger bundles to restrain wind effects on natural draft dry cooling system

    International Nuclear Information System (INIS)

    Liao, H.T.; Yang, L.J.; Du, X.Z.; Yang, Y.P.

    2016-01-01

    Highlights: • Triangularly arranged heat exchanger around the dry-cooling tower is proposed. • By coupling condenser with dry cooling system, TACHE performance is obtained. • At low wind speeds, cooling performance with TACHE is inferior to that with CACHE. • Better performance can be achieved for cooling system with TACHE at high wind speeds. • TACHE can be applied to the region with the strong prevailing wind all year around. - Abstract: It has been commonly recognized that the crosswind may deteriorate the cooling performance of the natural draft dry cooling system with vertically arranged heat exchanger bundles around the circumference of dry-cooling tower. With the purpose for restraining the adverse effects of ambient winds, a novel triangular configuration of heat exchanger bundles is proposed in this work. The air-side flow and heat transfer models coupled with the circulating water heat transfer process are developed for two kinds of natural draft dry cooling systems with the conventional circularly arranged and novel triangularly arranged heat exchanger bundles, by which the flow and temperature fields, mass flow rate of cooling air, outlet water temperature of heat exchanger and turbine back pressure are obtained. Three wind directions of 0°, 90°, and 180° are investigated at various wind speeds for the natural draft dry cooling system with triangularly arranged heat exchanger bundles, which are compared with the conventional system with circularly arranged heat exchanger bundles. The results show that the thermo-flow performances of the natural draft dry cooling system with triangularly arranged heat exchanger get improved significantly at high wind speeds and in the wind direction of 180°, thus a low turbine back pressure can be achieved, which is of benefit to the energy efficiency of the power generating unit. The natural draft dry cooling system with triangularly arranged heat exchanger is recommended to apply to the regions with

  17. Experimental measurements of static pressure and pressure drop in a duct enclosing a seven wire-wrapped rod bundle

    International Nuclear Information System (INIS)

    Graca, M.C.; Ballve, H.; Fernandez y Fernandez, E.; Carajilescov, P.

    1981-01-01

    The friction factor and the static pressure distributions, in the axial and transversal directions, in the wall of the hexagonal duct, enclosing a seven wire-wrapped rod bundle, were experimentally measured, using an air opened loop. The Reynolds numbers are the range 10 3 - 5x10 4 . The friction factors are compared to existing correlations. The static pressure distributions show that the static pressure is not hydrostatic in the cross section of the flow. (Author) [pt

  18. Use of a commercial heat transfer code to predict horizontally oriented spent fuel rod temperatures

    International Nuclear Information System (INIS)

    Wix, S.D.; Koski, J.A.

    1992-01-01

    Radioactive spent fuel assemblies are a source of hazardous waste that will have to be dealt with in the near future. It is anticipated that the spent fuel assemblies will be transported to disposal sites in spent fuel transportation casks. In order to design a reliable and safe transportation cask, the maximum cladding temperature of the spent fuel rod arrays must be calculated. The maximum rod temperature is a limiting factor in the amount of spent fuel that can be loaded in a transportation cask. The scope of this work is to demonstrate that reasonable and conservative spent fuel rod temperature predictions can be made using commercially available thermal analysis codes. The demonstration is accomplished by a comparison between numerical temperature predictions, with a commercially available thermal analysis code, and experimental temperature data for electrical rod heaters simulating a horizontally oriented spent fuel rod bundle

  19. 2D model for melt progression through rods and debris

    International Nuclear Information System (INIS)

    Fichot, F.

    2001-01-01

    During the degradation of a nuclear core in a severe accident scenario, the high temperatures reached lead to the melting of materials. The formation of liquid mixtures at various elevations is followed by the flow of molten materials through the core. Liquid mixture may flow under several configurations: axial relocation along the rods, horizontal motion over a plane surface such as the core support plate or a blockage of material, 2D relocation through a debris bed, etc.. The two-dimensional relocation of molten material through a porous debris bed, implemented for the simulation of late degradation phases, has opened a new way to the elaboration of the relocation model for the flow of liquid mixture along the rods. It is based on a volume averaging method, where wall friction and capillary effects are taken into account by introducing effective coefficients to characterize the solid matrix (rods, grids, debris, etc.). A local description of the liquid flow is necessary to derive the effective coefficients. Heat transfers are modelled in a similar way. The derivation of the conservation equations for the liquid mixture falling flow (momentum) in two directions (axial and radial-horizontal) and for the heat exchanges (energy) are the main points of this new model for simulating melt progression. In this presentation, the full model for the relocation and solidification of liquid materials through a rod bundle or a debris bed is described. It is implemented in the ICARE/CATHARE code, developed by IPSN in Cadarache. The main improvements and advantages of the new model are: A single formulation for liquid mixture relocation, in 2D, either through a rod bundle or a porous debris bed, Extensions to complex structures (grids, by-pass, etc..), The modeling of relocation of a liquid mixture over plane surfaces. (author)

  20. On the perfect hexagonal packing of rods

    International Nuclear Information System (INIS)

    Starostin, E L

    2006-01-01

    In most cases the hexagonal packing of fibrous structures or rods extremizes the energy of interaction between strands. If the strands are not straight, then it is still possible to form a perfect hexatic bundle. Conditions under which the perfect hexagonal packing of curved tubular structures may exist are formulated. Particular attention is given to closed or cycled arrangements of the rods like in the DNA toroids and spools. The closure or return constraints of the bundle result in an allowable group of automorphisms of the cross-sectional hexagonal lattice. The structure of this group is explored. Examples of open helical-like and closed toroidal-like bundles are presented. An expression for the elastic energy of a perfectly packed bundle of thin elastic rods is derived. The energy accounts for both the bending and torsional stiffnesses of the rods. It is shown that equilibria of the bundle correspond to solutions of a variational problem formulated for the curve representing the axis of the bundle. The functional involves a function of the squared curvature under the constraints on the total torsion and the length. The Euler-Lagrange equations are obtained in terms of curvature and torsion and due to the existence of the first integrals the problem is reduced to the quadrature. The three-dimensional shape of the bundle may be readily reconstructed by integration of the Ilyukhin-type equations in special cylindrical coordinates. The results are of universal nature and are applicable to various fibrous structures, in particular, to intramolecular liquid crystals formed by DNA condensed in toroids or packed inside the viral capsids

  1. Results of the QUENCH-12 experiment on reflood of a VVER-type bundle

    Energy Technology Data Exchange (ETDEWEB)

    Stuckert, J.; Grosse, M.; Heck, M.; Schanz, G.; Sepold, L.; Stegmaier, U.; Steinbrueck, M. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Materialforschung, Programm Nukleare Sicherheitsforschung; Goryachev, A.; Ivanova, I. [RIAR (FSUE SSC-RIAR) Dimitrovgrad (Russian Federation)

    2008-09-15

    The QUENCH experiments are to investigate the hydrogen source term resulting from the water injection into an uncovered core of a Light-Water Reactor. The QUENCH test bundle with a total length of approximately 2.5 m usually consists of 21 fuel rod simulators of Western PWR (Pressurized Water Reactor) geometry. The QUENCH-12 test bundle, however, which was set up to investigate the effects of VVER materials and bundle geometry (hexagonal lattice) on core reflood consisted of 31 fuel rod simulators. 18 rods of which were electrically heated using tungsten heaters in the rod center. All claddings, corner rods and grid spacers were made of Zr1%Nb (E110) and the shroud of Zr2.5%Nb (E125). For comparison, the QUENCH-06 test (ISP-45) with Western PWR geometry (square lattice) was chosen as reference. QUENCH-12 conducted at the Forschungszentrum Karlsruhe (FZK, Karlsruhe Research Center) on 27 September, 2006 in the frame of the EC-supported ISTC program 1648.2 was proposed by FZK together with RIAR Dimitrovgrad and IBRAE Moscow (Russia), and supported by pretest calculations performed by PSI (Switzerland) and the Kurchatov Institute Moscow (Russia) together with IRSN Cadarache (France). It had been preceded by a low-temperature (maximum 1073 K) pretest on 25 August, 2006 to characterize the bundle thermal hydraulic performance and to provide data to assess the code models used for pretest calculational support. After a stabilization period at 873 K pre-oxidation took place at {proportional_to}1470 K for {proportional_to}3400 s to achieve a maximum oxide thickness of about 200 {mu}m. A transient phase followed with a temperature rise to {proportional_to}2050 K. Then quenching of the bundle by a water flow of 48 g/s was initiated cooling the bundle to ambient temperature in {proportional_to}5 min. Following reflood initiation, a moderate temperature excursion of {proportional_to}50 K was observed, over a longer period than in QUENCH-06. The temperatures at elevations

  2. Results of the QUENCH-12 experiment on reflood of a VVER-type bundle

    International Nuclear Information System (INIS)

    Stuckert, J.; Grosse, M.; Heck, M.; Schanz, G.; Sepold, L.; Stegmaier, U.; Steinbrueck, M.

    2008-09-01

    The QUENCH experiments are to investigate the hydrogen source term resulting from the water injection into an uncovered core of a Light-Water Reactor. The QUENCH test bundle with a total length of approximately 2.5 m usually consists of 21 fuel rod simulators of Western PWR (Pressurized Water Reactor) geometry. The QUENCH-12 test bundle, however, which was set up to investigate the effects of VVER materials and bundle geometry (hexagonal lattice) on core reflood consisted of 31 fuel rod simulators. 18 rods of which were electrically heated using tungsten heaters in the rod center. All claddings, corner rods and grid spacers were made of Zr1%Nb (E110) and the shroud of Zr2.5%Nb (E125). For comparison, the QUENCH-06 test (ISP-45) with Western PWR geometry (square lattice) was chosen as reference. QUENCH-12 conducted at the Forschungszentrum Karlsruhe (FZK, Karlsruhe Research Center) on 27 September, 2006 in the frame of the EC-supported ISTC program 1648.2 was proposed by FZK together with RIAR Dimitrovgrad and IBRAE Moscow (Russia), and supported by pretest calculations performed by PSI (Switzerland) and the Kurchatov Institute Moscow (Russia) together with IRSN Cadarache (France). It had been preceded by a low-temperature (maximum 1073 K) pretest on 25 August, 2006 to characterize the bundle thermal hydraulic performance and to provide data to assess the code models used for pretest calculational support. After a stabilization period at 873 K pre-oxidation took place at ∝1470 K for ∝3400 s to achieve a maximum oxide thickness of about 200 μm. A transient phase followed with a temperature rise to ∝2050 K. Then quenching of the bundle by a water flow of 48 g/s was initiated cooling the bundle to ambient temperature in ∝5 min. Following reflood initiation, a moderate temperature excursion of ∝50 K was observed, over a longer period than in QUENCH-06. The temperatures at elevations between 850 mm and 1050 mm exceeded the melting temperature of β-Zr, i

  3. Critical heat flux and post-critical heat flux performance of a 6-m, 37-element fully segmented bundle cooled by Freon-12

    International Nuclear Information System (INIS)

    Nickerson, J.R.

    1982-05-01

    A 6-m, 37-element, electrically heated bundle with full end plate simulation, cooled by Freon-12, has been tested for CHF (critical heat flux) and post-CHF conditions in the MR-3 Freon loop. The bundle was tested in a horizontal attitude and had a uniform axial heat flux distribution and radial heat flux depression. A total of 110 CHF points have been collected over the following range of water equivalent conditions: exit pressure 8.27 - 11.03 MPa, mass flux 1.38 - 8.14 Mg.m -2 .s -1 , inlet subcooling 0 - 500 kJ.kg -1 , outlet quality 10% - 37%. The data have been correlated on both a systems and local conditions basis over a limited mass flux range to within 2.8% rms. Significant CHF increases over smooth bundle results have been observed along with significant CHF improvement over a two end plate bundle simulation in the lower mass flux ranges. A satisfactory axial drypatch spreading correlation has been determined and extensive drypatch wall superheat mapping has been performed

  4. Dimensional measurement of fresh fuel bundle for CANDU reactor

    International Nuclear Information System (INIS)

    Jo, Chang Keun; Cho, Moon Sung; Suk, Ho Chun; Koo, Dae Seo; Jun, Ji Su; Jung, Jong Yeob

    2005-01-01

    This report describes the results of the dimensional measurement of fresh fuel bundles for the CANDU reactor in order to estimate the integrity of fuel bundle in two-phase flow in the CANDU-6 fuel channel. The dimensional measurements of fuel bundles are performed by using the 'CANDU Fuel In-Bay Inspection and Dimensional Measurement System', which was developed by this project. The dimensional measurements are done from February 2004 to March 2004 in the CANDU fuel storage of KNFC for the 36 fresh fuel bundles, which are produced by KNFC and are waiting for the delivery to the Wolsong-3 plant. The detail items of dimensional measurements are included fuel rod and bearing pad profiles of the outer ring in fuel bundle, diameter of fuel bundle, bowing of fuel bundle, fuel rod length, and surface profile of end plate profile. The measurement data will be compared with those of the post-irradiated bundles cooled in Wolsong-3 NPP spent fuel pool by using the same bundles and In-Bay Measurement System. So, this analysis of data will be applied for the evaluation of fuel bundle integrity in two-phase flow of the CANDU-6 fuel channel

  5. Recent High Heat Flux Tests on W-Rod-Armored Mockups

    International Nuclear Information System (INIS)

    Nygren, Richard E.; Youchison, Dennis L.; McDonald, Jimmie M.; Lutz, Thomas J.; Miszkiel, Mark E.

    2000-01-01

    In the authors initial high heat flux tests on small mockups armored with W rods, done in the small electron beam facility (EBTS) at Sandia National Laboratories, the mockups exhibited excellent thermal performance. However, to reach high heat fluxes, they reduced the heated area to only a portion (approximately25%) of the sample. They have now begun tests in their larger electron beam facility, EB 1200, where the available power (1.2 MW) is more than enough to heat the entire surface area of the small mockups. The initial results indicate that, at a given power, the surface temperatures of rods in the EB 1200 tests is somewhat higher than was observed in the EBTS tests. Also, it appears that one mockup (PW-10) has higher surface temperatures than other mockups with similar height (10mm) W rods, and that the previously reported values of absorbed heat flux on this mockup were too high. In the tests in EB 1200 of a second mockup, PW-4, absorbed heat fluxes of approximately22MW/m 2 were reached but the corresponding surface temperatures were somewhat higher than in EBTS. A further conclusion is that the simple 1-D model initially used in evaluating some of the results from the EBTS testing was not adequate, and 3-D thermal modeling will be needed to interpret the results

  6. Burnout experiments with 6 x 6, 8 x 8 and 7 x 7 rod bundle test sections using freon as model fluid

    International Nuclear Information System (INIS)

    Fulfs, H.; Katsaounis, A.; Minden, C.v.

    1976-01-01

    This paper reports on burnout experiments at staedy state condition using Freon12 as model fluid. The experiments were carried out with three test sections with 6 x 6, 8 x 8 and 7 x 7 rod bundles. The axial flux distribution of the rods is either constant or reactor like. The transformed measured points using STEVENS and BOURE scaling factors to equivalent water conditions respectively, were compared to the burnout correlation W3 using the reactor layout program DYNAMIT. The DYNAMIT code is a thermohydraulic lay-out reactor program without consideration of mixing flow between the subchannels. (orig.) [de

  7. An assessment of the critical heat flux approaches of thermal-hydraulic system analysis codes using bundle data from the Heat Transfer Research Facility

    International Nuclear Information System (INIS)

    Min Lee

    1994-01-01

    Critical heat flux (CHF) bundle data from the Heat Transfer Research Facility of Columbia University are used to check the validity of the CHF approaches used in thermal-hydraulic system analysis codes for light water reactors. The CHF approaches assessed include the Biasi et al. correlation of TRAC, the Groeneveld et al. CHF table lookup approach of RELAP5/MOD3, the CHF table lookup approach of CATHARE, and the CHF approach of RETRAN. Depending on system pressure, RETRAN uses the B and W2, Barnett, and modified Barnett correlations and a linear interpolation scheme to predict CHF. Results show that among these CHF approaches, the Groeneveld et al. approach has the best prediction accuracy and the smallest uncertainty in the estimation of the HTRF bundle data. On the average, the Groeneveld et al. approach overpredicts the uniform axial heat flux distribution by 3.6% and the nonuniform axial heat flux distribution by 0.9%. The performance of the RETRAN approach is comparable with that of the Groenevel et al. Approach for uniform axial heat flux. In general, the accuracy and the uncertainty of all the approaches, except that of CATHARE, are worse under a nonuniform axial heat distribution than under a uniform axial heat distribution. All the CHF approaches assessed have a tendency to overpredict the HTRF bundle data at low pressure, low measured CHF, and high CHF quality. The performance of the Groenevel et al. approach is improved through a CHF table update and modification of the bundle correction factor using the HTRF bundle data

  8. Effects of Angle of Rotation on Pool Boiling Heat Transfer of V-shape Tube Bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Myeong-Gie [Andong National University, Andong (Korea, Republic of)

    2016-10-15

    The most important facility for the systems is a passive heat exchanger that transfers core decay heat to the cold water in a water storage tank under atmospheric pressure. Since the space for the installation of the heat exchanger is usually limited, developing more efficient heat exchangers is important. In general, pool boiling is generated on the surface of the heat exchanging tube. The major design parameter of the heat exchanger is a tube arrangement. The upper tube is affected by the lower tube and the enhancement of the heat transfer on the upper tube is estimated by the bundle effect. Since heat transfer is related to the conditions of a tube surface, bundle geometries, and a liquid type, lots of studies have been carried out for the combinations of those parameters. An experimental study was performed to investigate the effects of the angle of rotation on pool boiling heat transfer of a V-shape tube bundle. For the test, two smooth stainless steel tubes of 19 mm outside diameter and the water at atmospheric pressure were used. The enhancement of the heat transfer is clearly observed when the angle becomes to 90° where the upper tube has the maximum region of influence by the lower tube. The convective flow and liquid agitation enhance heat transfer while the coalescence of the bubbles deteriorates heat transfer.

  9. BWR fuel assembly with improved spacer and fuel bundle design for enhanced thermal-hydraulic performance

    International Nuclear Information System (INIS)

    Mildrum, C.M.; Taleyarkhan, R.P.

    1987-01-01

    In a fuel assembly having a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods, an outer tubular flow channel surrounding the fuel rods so as to direct flow of coolant/moderator fluid along the fuel rods, a hollow water cross extending centrally through and interconnected with the outer flow channel so as to divide the channel into separate compartments and the bundle of fuelrods into a plurality of mini-bundles thereof being disposed in the compartments, and spacers axially displaced along the fuel rods in each of the mini-bundles thereof. Each spacer is composed of inner and outer means which together define spacer cells at corner, side and interior locations of the spacer and have respective protrusions formed thereon which extend into cells so as to maintain the fuel rods received through the spacer cells in laterally spaced relationships. The improvement is described which comprises: (a) a generally uniform poison coating within at least a majority of the fuel rods; (b) a predetermined pattern of fuel enrichment with respect to the fuel rods of each mini-bundle thereof which together with the uniform poison coating within the fuel rods ensures that the packing powers of the fuel rods in the corner and side cells of the spacers are less than the peaking power of a leading one of the fuel rods in the interior cells of the spacers; and (c) each of the fuel rods being received through the cells of each spacer having a diametric size smaller than that of each of the fuel rods received through the side and interior cells of each spacer, the diametric sizes of each of the fuel rods received through the side and interior cells of each spacer being generally equal

  10. Experience using individually supplied heater rods in critical power testing of advanced BWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Majed, M.; Morback, G.; Wiman, P. [ABB Atom AB, Vasteras (Sweden)] [and others

    1995-09-01

    The ABB Atom FRIGG loop located in Vasteras Sweden has during the last six years given a large experience of critical power measurements for BWR fuel designs using indirectly heated rods with individual power supply. The loop was built in the sixties and designed for maximum 100 bar pressure. Testing up to the mid eighties was performed with directly heated rods using a 9 MW, 80 kA power supply. Providing test data to develop critical power correlations for BWR fuel assemblies requires testing with many radial power distributions over the full range of hydraulic conditions. Indirectly heated rods give large advantages for the testing procedure, particularly convenient for variation of individual rod power. A test method being used at Stern Laboratories (formerly Westinghouse Canada) since the early sixties, allows one fuel assembly to simulate all required radial power distributions. This technique requires reliable indirectly heated rods with independently controlled power supplies and uses insulated electric fuel rod simulators with built-in instrumentation. The FRIGG loop was adapted to this system in 1987. A 4MW power supply with 10 individual units was then installed, and has since been used for testing 24 and 25 rod bundles simulating one subbundle of SVEA-96/100 type fuel assemblies. The experience with the system is very good, as being presented, and it is selected also for a planned upgrading of the facility to 15 MW.

  11. Thermal-hydraulic performance analysis of a subchannel with square and triangle fuel rod arrangements using the entropy generation approach

    Institute of Scientific and Technical Information of China (English)

    S.Talebi; M.M.Valoujerdi

    2017-01-01

    The present paper discusses entropy generation in fully developed turbulent flows through a subchannel,arranged in square and triangle arrays.Entropy generation is due to contribution of both heat transfer and pressure drop.Our main objective is to study the effect of key parameters such as spacer grid,fuel rod power distribution,Reynolds number Re,dimensionless heat power ω,lengthto-fuel-diameter ratio λ,and pitch-to-diameter ratio ξ on subchannel entropy generation.The analysis explicitly shows the contribution of heat transfer and pressure drop to the total entropy generation.An analytical formulation is introduced to total entropy generation for situations with uniform and sinusoidal rod power distribution.It is concluded that power distribution affects entropy generation.A smoother power profile leads to less entropy generation.The entropy generation of square rod array bundles is more efficient than that of triangular rod arrays,and spacer grids generate more entropy.

  12. A comprehensive in-pile test of PWR fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kang Rixin; Zhang Shucheng; Chen Dianshan (Academia Sinica, Beijing (China). Inst. of Atomic Energy)

    1991-02-01

    An in-pile test of PWR fuel bundle has been conducted in HWRR at IAE of China. This paper describes the structure of the test bundle (3x3-2), fabrication process and quality control of the fuel rod, irradiation conditions and the main Post Irradiation Examination (PIE) results. The test fuel bundle was irradiated under the PWR operation and water chemistry conditions with an average linear power of 381 W/cm and reached an average burnup of 25010 MWd/tU of the fuel bundle. After the test, destructive and non-destructive examination of the fuel rods was conducted at hot laboratories. The fission gas release was 10.4-23%. The ridge height of cladding was 3 to 8 {mu}m. The hydrogen content of the cladding was 80 to 140 ppm. The fuel stack height was increased by 2.9 to 3.3 mm. The relative irradiation growth was about 0.11 to 0.17% of the fuel rod length. During the irradiation test, no fuel rod failure or other abnormal phenomena had been found by the on-line fuel failure monitoring system of the test loop and water sampling analysis. The structure of the test fuel assembly was left undamaged without twist and detectable deformation. (orig.).

  13. Experimental study on the convective heat transfer enhancement in single-phase steam flow by a support grid

    International Nuclear Information System (INIS)

    Kim, Byoung Jae; Kim, Kihwan; Kim, Dong-Eok; Youn, Young-Jung; Park, Jong-Kuk; Moon, Sang-Ki; Song, Chul-Hwa

    2014-01-01

    Highlights: • The convective heat transfer enhancement by support grids is investigated. • Experiments were performed in a square array 2 × 2 rod bundle. • The enhancement was affected not only by the blockage ratio also by the Reynolds number. • For low Reynolds numbers, the enhancement depends on the Reynolds number (Re). • For high Reynolds numbers, the enhancement is nearly independent of Re. - Abstract: Single-phase flow occurs in the fuel rod bundle of a pressurized water reactor, during the normal operation period or at the early stage of the reflood phase in a loss-of-coolant accident scenario. In the former period, the flow is single-phase water flow, but in the latter case, the flow is single-phase steam flow. Support grids are required to maintain a proper geometry configuration of fuel rods within nuclear fuel assemblies. This study was conducted to elucidate the effects of support grids on the convective heat transfer in single-phase steam flow. Experiments were made in a square array 2 × 2 rod bundle. The four electrically-heating rods were maintained by support grids with mixing vanes creating a swirl flow. Two types of support grids were considered in this study. The two types are geometrically similar except the blockage ratio by different mixing vane angles. For all test runs, 2 kW power was supplied to each rod. The working fluid was superheated steam with Re = 2,301–39,594. The axial profile of the rod surface temperatures was measured, and the convective heat transfer enhancement by the presence of the support grids was examined. The peak heat transfer enhancement was a function of not only the blockage ratio but also the Reynolds number. Given the same blockage ratio, the heat transfer enhancement was sensitive to the Reynolds number in laminar flow, whereas it was nearly independent of the Reynolds number in turbulent flow

  14. The button effect of CANFLEX bundle on the critical heat flux and critical channel power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Jisu; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G R; Bullock, D E; Inch, W [Atomic Energy of Canada Limited, Ontario (Canada)

    1998-12-31

    A CANFLEX (CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5 mm) to reduce maximum element power rating and buttons to enhance the critical heat flux (CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant R-134a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power (CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF`s for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only for high pressure conditions. It indicates that the button is a useful design for CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions. 5 refs., 11 figs. (Author)

  15. The button effect of CANFLEX bundle on the critical heat flux and critical channel power

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Jun, Jisu; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Dimmick, G. R.; Bullock, D. E.; Inch, W. [Atomic Energy of Canada Limited, Ontario (Canada)

    1997-12-31

    A CANFLEX (CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5 mm) to reduce maximum element power rating and buttons to enhance the critical heat flux (CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant R-134a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power (CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF`s for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only for high pressure conditions. It indicates that the button is a useful design for CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions. 5 refs., 11 figs. (Author)

  16. Fuel-clad heat transfer coefficient of a defected fuel rod

    International Nuclear Information System (INIS)

    Bruet, M.; Stora, J.P.

    1976-01-01

    A special rod has been built with a stack of UO 2 pellets inside a thick zircaloy clad. The atmosphere inside the fuel rod can be changed and particularly the introduction of water is possible. The capsule was inserted in the Siloe pool reactor in a special device equipped with a neutron flux monitor. The fuel centerline temperature and the temperature at a certain radius of the clad were recorded by two thermocouples. The temperature profiles in the fuel and in the cladding have been calculated and then the heat transfer coefficient. In order to check the proper functioning of the device, two runs were successively achieved with a helium atmosphere. Then the helium atmosphere inside the fuel rod was removed and replaced by water. The heat transfer coefficients derived from the measurements at low power level are in agreement with the values given by the model based on thermal conductivity. However, for higher power levels, the heat transfer coefficients become higher than those based on the calculated gap

  17. Heat transfer and friction on smooth and rough test rods

    International Nuclear Information System (INIS)

    Franken, W.M.P.; Hoogland, H.; Deijman, P.

    1977-06-01

    Results are reported on heat transfer and pressure drop tests on one smooth and nine rough test rods in an annular geometry. The wall roughness consisted of transversal ribs with various roughness pitches, rib heights and rib widths. The tests were performed with air as coolant under a wide range of experimental conditions: 10 5 5 , 1.1 2. Special attention has been given to the effect of variation of the physical coolant properties over the flow cross section. This effect could be described by the power function (Tsub(w)/Tsub(b))sup(-0.3l) in additional systematic variation of the heat transfer could be recognized, dependent on the coolant temperature level. The experimental results were correlated by the equation St = C(Tsub(in)) Resup(-0.2) Prsup(-0.6) (Tsub(w)/Tsub(b)sup(-0.31). Values of C(Tsub(in)) are given in tabular form. The thermal entrance effect has been measured on various test rods. A substantial reduction of the heat transfer coefficient was almost constant along the rough test rods

  18. CFD in supercritical water-cooled nuclear reactor (SCWR) with horizontal tube bundles

    International Nuclear Information System (INIS)

    Shang, Zhi; Lo, Simon

    2009-01-01

    The commercial CFD code STAR-CD 4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round tubes and tube bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal tubes and tube bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. From the study of single round tubes, the Speziale quadratic non-linear high-Re k-ε turbulence model with the two-layer model for near wall treatment is found to produce the best results in comparison with experimental data. In tube bundle simulations, it is found that the temperature is higher in the top half of the bundle and the highest tube wall temperature is located at the outside tubes where the flow rate is the lowest. The secondary flows across the bundle are highly complex. Their main effect is to even out the temperature over the area within each individual recirculating region. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR. (author)

  19. CFD in supercritical water-cooled nuclear reactor (SCWR) with horizontal tube bundles

    Energy Technology Data Exchange (ETDEWEB)

    Zhi Shang, E-mail: zhi.shang@stfc.ac.uk [Science and Technology Facilities Council, Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Lo, Simon, E-mail: simon.lo@uk.cd-adapco.com [CD-adapco, Trident House, Basil Hill Road, Didcot OX11 7HJ (United Kingdom)

    2011-11-15

    The commercial CFD code STAR-CD 4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round tubes and tube bundles. Reactors with vertical or horizontal flow in the core can be found. In a vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in a horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal tubes and tube bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. From the study of single round tubes, the Speziale quadratic non-linear high-Re k-{epsilon} turbulence model with the two-layer model for near wall treatment is found to produce the best results in comparison with experimental data. In tube bundle simulations, it is found that the temperature is higher in the top half of the bundle and the highest tube wall temperature is located at the outside tubes where the flow rate is the lowest. The secondary flows across the bundle are highly complex. Their main effect is to even out the temperature over the area within each individual recirculation region. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR.

  20. Maximum allowable heat flux for a submerged horizontal tube bundle

    International Nuclear Information System (INIS)

    McEligot, D.M.

    1995-01-01

    For application to industrial heating of large pools by immersed heat exchangers, the socalled maximum allowable (or open-quotes criticalclose quotes) heat flux is studied for unconfined tube bundles aligned horizontally in a pool without forced flow. In general, we are considering boiling after the pool reaches its saturation temperature rather than sub-cooled pool boiling which should occur during early stages of transient operation. A combination of literature review and simple approximate analysis has been used. To date our main conclusion is that estimates of q inch chf are highly uncertain for this configuration

  1. Visual observations of a degraded bundle of irradiated fuel: the Phebus FPT1 test

    International Nuclear Information System (INIS)

    Barrachin, M.; Bottomley, P.D.

    1999-01-01

    The international Phebus-FP (Fission Product) project is managed by the Institut de Protection et Surete Nucleaire in collaboration with Electricite de France (EDF), the European Commission (EC), the USNRC (USA), COG (Canada), NUPEC and JAERI (Japan), KAERI (South Korea), PSI and HSK (Switzerland). It is designed to measure the source-term and to study the degradation of irradiated UO 2 fuel in conditions typical of a severe loss of coolant accident in a pressurised water reactor (PWR). In the first test (FPT0), performed in December '93, a bundle of 20 fresh fuel rods and a central Ag-In-Cd control rod underwent a short 15-day irradiation to generate fission products before testing in the Phebus reactor in Cadarache. The second test (FPT1) was performed in July '96, in the same conditions and geometry, but using irradiated fuel (-23 GWd/tU). In the FPT1 test, the bundle was heated to an estimated 3000 K over a period of 30 minutes in order to induce a substantial liquefaction of the bundle. After the test, the bundle was embedded in epoxy and cut at different levels to investigate the mechanisms of the core degradation. This paper reports the visual observations of the degraded FPT1 bundle, very preliminary interpretations about the scenario of degradation and a comparison between the behaviour of the fuel in the FPT0 and FPT1 tests. (author)

  2. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    Long, J.W.; Flora, B.S.; Ford, K.L.

    1980-01-01

    The invention relates to a nuclear power reactor fuel bundle of the type wherein several rods are mounted in parallel array between two tie plates which secure the fuel rods in place and are maintained in assembled position by means of a number of tie rods secured to both of the end plates. Improved apparatus is provided for attaching the tie rods to the upper tie plate by the use of locking lugs fixed to rotatable sleeves which engage the upper tie plate. (auth)

  3. Thermal hydraulic test of advanced fuel bundle with spectral shift rod (SSR) for BWR. Effect of thermal hydraulic parameters on steady state characteristics

    International Nuclear Information System (INIS)

    Kondo, Takao; Kitou, Kazuaki; Chaki, Masao; Ohga, Yukiharu; Makigami, Takeshi

    2011-01-01

    Japanese national project of next generation light water reactor (LWR) development started in 2008. Under this project, spectral shift rod (SSR) is being developed. SSR, which replaces conventional water rod (WR) of boiling water reactor (BWR) fuel bundle, was invented to enhance the BWR's merit, spectral shift effect for uranium saving. In SSR, water boils by neutron and gamma-ray direct heating and water level is formed as a boundary of the upper steam region and the lower water region. This SSR water level can be controlled by core flow rate, which amplifies the change of average core void fraction, resulting in the amplified spectral shift effect. This paper presents the steady state test results of the base geometry case in SSR thermal hydraulic test, which was conducted under the national project of next generation LWR. In the test, thermal hydraulic parameters, such as flow rate, pressure, inlet subcooling and heater rod power are changed to evaluate these effects on SSR water level and other SSR characteristics. In the test results, SSR water level rose as flow rate rose, which showed controllability of SSR water level by flow rate. The sensitivities of other thermal hydraulic parameters on SSR water level were also evaluated. The obtained data of parameter's sensitivities is various enough for the further analytical evaluation. The fluctuation of SSR water level was also measured to be small enough. As a result, it was confirmed that SSR's steady state performance was as planned and that SSR design concept is feasible. (author)

  4. CFD method research on characteristics cells in rod bundle fuel assembly

    International Nuclear Information System (INIS)

    Chen Jie; Chen Bingyan; Zhang Hong

    2011-01-01

    Two characteristic cells are in AFA-3G fuel assembly, that is typical cell and control rod guide cell. And there are some rules on the arrangement of mixing vanes. For the two characteristic cells, mixing capability is evaluated axially from the point of the first and second kind of sub-channel with CFD method. Mass mixing and heat mixing are interaction but different with each other. Although the mass mixing in the first kind of sub-channel is stronger, the thermal capability of the two is to some tune from the point of heat transfer. In the experiment research on thermal-hydraulic performance of AFA-3G fuel assembly, the arrangements of mixing vanes should refer to the two spacer grids of characteristic cells. (authors)

  5. Computation of 3D thermohydraulics in partially blocked bundles during the reflood phase of a LOCA

    International Nuclear Information System (INIS)

    Cicero, G.M.; Briere, E.; Fornaciari, G.

    1994-06-01

    In Pressurized Water Reactors (PWR), ballooning of the fuel rod claddings may occur during a LOCA, since the fuel rod claddings are heated up, and the system pressure is low. The severe blockages that may result induce cross-flow diversion and three-dimensional effects on thermohydraulics in the core bundle, during the reflood phase. To improve the knowledge of these phenomena and their physical modelling in the code CATHARE, 3D computer codes are needed. In 1990, EDF has started up a development and validation program of the 3D THYC computer code to analyze the thermohydraulics of the flow during the reflood phase, in partially blocked bundles. The main objective is to calculate the temperatures of the rods above the quench front, when they are cooled by superheated steam with saturated droplets. First, this paper introduces the THYC model developed for reflood studies. Secondly, we report the first qualification results on a Flooding Experiments with Blocked Array (FEBA) test. Thirdly, we analyze the model predictions on a large break LOCA transient, in a 900 MW PWR 11x11 core area with a 3x3 central blockage. THYC simulates the transient in the bundle around and above the blockage, until the quench front enters the computational domain. Previously, a 1D CATHARE simulation gives the boundary conditions and, in the reactor core case, the deformation of the blocked fuel rods. The results analysis focused on the time evolution of the clad temperatures in the blocked and in the bypass region. In the FEBA test simulation, the main observations are properly predicted within the blockage. Temperatures are lower in blocked rod sleeves than in unblocked rod claddings since the steam gap reduces the power transmitted by the heater rod to the sleeve. In the core case, the model predicts the opposite result. Within the blockage, ballooned rod temperatures are higher than non-ballooned rod ones. We show by sensitivity studies that these behaviour difference between FEBA rods

  6. Freely suspended rod fall dampener, especially for control rod of liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Becvar, J.; Saroch, V.

    1977-01-01

    A shock absorber is described whose advantage is that the space required for the movement of the shock absorber in the operating travel of the system suspension rod-control rod bundle may be reduced. The design allows the automatic disconnection of the system and the removal of the suspension rod from the reactor without dismantling. The braking force reaction is transmitted to the structure above the core. The system fall energy is absorbed on the side of the suspension rod which has a bigger mass. (J.B.)

  7. Study of the pressure drop in a rob bundle arranged in square array

    International Nuclear Information System (INIS)

    Barros Filho, J.A.

    1987-01-01

    Results of single-phase tests performed in a 3 x 3 rod bundle arranged in square array are presented and analysed. The tests were performed in adiabatic conditions and with heat transfer, covering the following range of parameters: Reynolds no.: 1,5 to 20 x 10 4 ; inlet temperature [ 0 C]: 30 to 150; pressure [bar]: 1 to 15; heat flux [kW/cm 2 ]: 0 to 1000. Correlations were determined for the friction factor, isothermal under conditions of heat transfer, spacer grids pressure drop coefficient and average heat transfer coefficient. The experimental data were compared with published data obtained by other researchers and with some theoretical models selected in the literature. (Autor) [pt

  8. Heat removal in gas-cooled fuel rod clusters

    International Nuclear Information System (INIS)

    Rehme, K.

    1975-01-01

    For a thermo- and fluid-dynamic analysis of fuel rod cluster subchannels for gas-cooled breeder reactors, the following values must be verified: a) friction coefficient as flow parameter; b) Stanton number as heat transfer parameter; c) influence of spacers on friction coefficient and Stanton number; d) heat and mass exchange between subchannels with different temperatures. These parameters are established by combining results of single experiments and of integral experiments. Mention is made of further studies to be performed in order to determine the heat removal from gas-cooled fast breeder fuel elements. (HR) [de

  9. Sensitivity Analysis of Gap Conductance for Heat Split in an Annular Fuel Rod

    International Nuclear Information System (INIS)

    Chun, Kun Ho; Chun, Tae Hyun; In, Wang Kee; Song, Keun Woo

    2006-01-01

    To increase of the core power density in the current PWR cores, an annular fuel rod was proposed by MIT. This annular fuel rod has two coolant channels and two cladding-pellet gaps unlike the current solid fuel rod. It's important to predict the heat split reasonably because it affects coolant enthalpy rise in each channel and Departure from Nuclear Boiling Ratio (DNBR) in each channel. Conversely, coolant conditions affect fuel temperature and heat split. In particular if the heat rate leans to either inner or outer channel, it is out of a thermal equilibrium. To control a thermal imbalance, placing another gap in the pellet is introduced. The heat flow distribution between internal and external channels as well as fuel and cladding temperature profiles is calculated with and without the fuel gap between the inner and outer pellets

  10. On the development of a grid-enhanced single-phase convective heat transfer correlation

    International Nuclear Information System (INIS)

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

    2011-01-01

    A new single-phase convective heat transfer augmentation correlation has been developed using single phase steam cooling experimental data obtained from the Penn State/NRC Rod Bundle Heat Transfer (RBHT) facility. Experimental data obtained from the RBHT single phase steam cooling tests have been evaluated and new findings identified. Previous rod bundle tests showed the importance of spacer grid on the local heat transfer, and that the augmentation in heat transfer downstream of a grid decays exponentially. The RBHT data also shows that the Reynolds number affects the rate at which this augmentation decays. The new correlation includes the strong dependence of heat transfer on both the Reynolds number and the grid blockage ratio. While the effects of both parameters were clearly evident in the RBHT experimental data, existing correlations do not account for the Reynolds number effect. The developed correlation incorporates Reynolds number in the decay curve of heat transfer. The newly developed correlation adequately accounts for the dependence of the heat transfer augmentation decay rate on the local flow Reynolds number. (author)

  11. Experiments on the fluid dynamics and thermodynamics of rod bundles to verify and support the design of SNR-300 fuel elements - status and open problems

    International Nuclear Information System (INIS)

    Moeller, R.; Weinberg, D.; Trippe, G.; Tschoeke, H.

    1978-01-01

    The reliable design of reactor core elements calls for precise knowledge of the 3D-temperature fields of the different components; this primarily applies to the fuel element cladding tubes, these being the first safety barrier. This paper describes and discusses where and how the 3D-temperature fields so far determined exclusively with the help of global thermohydraulic computer codes (SUBCHANNEL-Codes) have to be determined more accurately by local investigations. The basis of these investigations is the measurement of local velocities and temperatures in 19-rod bundle models of the SNR-300 fuel element performed at the Kernforschungszentrum Karlsruhe (KfK). Some important results of the extensive experimental investigations are reported and compared with global and local recalculations. Open problems are pointed out. The influence of the uncertainties in the thermohydraulic design with respect to the strength analysis are discussed. The most significant results and conclusions are: (1) The peripheral bundle region is the critical zone, which has to be investigated with priority. Here the maximal azimuthal temperature differences of the claddings are ten times higher than those in the central bundle region. (2) The present deviations between thermal experiments and global as well as local calculations are much too high. Within the parameters investigated a careful code adaptation to the experiments is of high priority. (3) The knowledge gaps concerning liquid metal heat transfer in irregular geometries have to be closed. (4) The hot-channel analysis has to be checked with respect to the latest more detailed knowledge of thermohydraulics. (author)

  12. Heat transfer to a dispersed two phase flow and detailed quench front velocity research

    International Nuclear Information System (INIS)

    De Boer, T.C.; Van der Molen, S.B.

    1985-01-01

    During the blow-down phase of a loss-off coolant accident (LOCA) in a pressurized water reactor the core will heat up dramatically. Water will be injected in the system, and by bottom flooding the core will be cooled. The use of one-dimensional computer models for the calculation of the reflood process in a bundle needs a better justification. The influence of an unheated shroud on prequench heat transfer is investigated in a tube, an annulus and a 4 rod bundle. By using a glass shroud for the annulus, optical analysis of the dispersed two-phase flow regime has been performed. The ECN 36-rod bundle tests as performed with axial uniform power profile are reflood and boil-down at 0.2 MPa pressure executed for different conditions. The experiment yield a data base suitable for code validation and development. Better understanding is obtained for the influence of the radial non-uniform temperature and/or power distributions on the reflood process. Heat transfer improvement induced by the presence of spacer grids is observed. 72 refs.; 220 figs.

  13. A study of critical heat flux in the fuel assembly dummies with various types of mixing grids

    International Nuclear Information System (INIS)

    Bezrukov, Yu. A.; Lisenkov, E. A.; Astakhov, V. I.; Vasilchenko, I. N.

    2013-01-01

    The report deals with the results of a study The report deals with the results of a study of critical heat fluxes (CHF) on the rod bundles equipped with mixing grids – intensifier of heat exchange. The study was carried out on FA dummies equipped with various kinds of mixing grids: “Vikhr” and “Sector screw die”. The tests were carried out at the following parameters: pressure 14 and 16 MPa; mass velocity 2000, 3000 and 4000 kg/(m2⋅s); relative enthalpy from minus 0,07 up to 0,18. Bundles with uniform and non-uniform axial heat generation were tested. The results of the experiments are intended for the elaboration of a correlation to be used in design development.of critical heat fluxes (CHF) on the rod bundles equipped with mixing grids – intensifier of heat exchange. The study was carried out on FA dummies equipped with various kinds of mixing grids: “Vikhr” and “Sector screw die”. The tests were carried out at the following parameters: pressure 14 and 16 MPa; mass velocity 2000, 3000 and 4000 kg/(m 2 ⋅s); relative enthalpy from minus 0,07 up to 0,18. Bundles with uniform and non-uniform axial heat generation were tested. The results of the experiments are intended for the elaboration of a correlation to be used in design development. (authors)

  14. Turbulent heat transfer to longitudinal flow through a triangular array of circular rods

    International Nuclear Information System (INIS)

    Pfann, J.

    1975-01-01

    Temperature distribution and heat transfer to longitudinal turbulent, fully developed flow through triangular arrays of smooth circular rods are analysed for liquids with Prandtl number approximately 1 and << 1. Nusselt number is plotted versus pitch and turbulence for constant heat flow and for constant temperature on the rod surface, and the optimum pitch is determined. The influence of Prandtl number is analysed. (Auth.)

  15. Study on Natural Convection around a vertical heated rod using PIV/LIF technique,

    International Nuclear Information System (INIS)

    Szijarto, R.; Yamaji, B.; Aszodi, A.

    2010-01-01

    The Nuclear Training Reactor of the Institute of Nuclear Techniques (Budapest University of Technology and Economics, Hungary) is a pool-type reactor with light water moderator and with a maximum thermal power of 100 kW. The fuel elements are cooled by natural convection. An experimental setup was built to analyse the nature of the natural convection around a heated rod. The flow field was investigated using an electrically heated rod, which models the geometry of a fuel pin in the training reactor. The electric power of the model rod is variable between 0-500 W. The rod was placed in a square-based glass tank. Particle Image Velocimetry and Laser Induced Fluorescence measurement techniques were used to study the velocity and temperature field in a two-dimensional area. The thermal and the hydraulic boundary layers were detected near a rod in a lower section of the aquarium. The laminar-turbulent transition of the flow regime was observed, the maximum velocity of the up-flow was 0.025-0.05 m/s. From the temperature measurements the local heat transfer coefficient was estimated. (Authors)

  16. NCEL: two dimensional finite element code for steady-state temperature distribution in seven rod-bundle

    International Nuclear Information System (INIS)

    Hrehor, M.

    1979-01-01

    The paper deals with an application of the finite element method to the heat transfer study in seven-pin models of LMFBR fuel subassembly. The developed code NCEL solves two-dimensional steady state heat conduction equation in the whole subassembly model cross-section and enebles to perform the analysis of thermal behaviour in both normal and accidental operational conditions as eccentricity of the central rod or full or partial (porous) blockage of some part of the cross-flow area. The heat removal is simulated by heat sinks in coolant under conditions of subchannels slug flow approximation

  17. Damping in heat exchanger tube bundles. A review

    International Nuclear Information System (INIS)

    Iqbal, Qamar; Khushnood, Shahab; Ghalban, Ali Roheim El; Sheikh, Nadeem Ahmed; Malik, Muhammad Afzaal; Arastu, Asif

    2007-01-01

    Damping is a major concern in the design and operation of tube bundles with loosely supported tubes in baffles for process shell and tube heat exchangers and steam generators which are used in nuclear, process and power generation industries. System damping has a strong influence on the amplitude of vibration. Damping depends upon the mechanical properties of the tube material, geometry of intermediate supports and the physical properties of shell-side fluid. Type of tube motion, number of supports, tube frequency, vibration amplitude, tube mass or diameter, side loads, support thickness, higher modes, shell-side temperature etc., affect damping in tube bundles. The importance of damping is further highlighted due to current trend of larger exchangers with increased shell-side velocities in modern units. Various damping mechanisms have been identified (Friction damping, Viscous damping, Squeeze film damping, Support damping. Two-Phase damping, and very recent-Thermal damping), which affect the performance of process exchangers and steam generators with respect to flow induced vibration design, including standard design guidelines. Damping in two-phase flow is very complex and highly void fraction, and flow-regime dependent. The current paper focuses on the various known damping mechanisms subjected to both single and two-phase cross-flow in process heat exchangers and steam generators and formulates the design guidelines for safer design. (author)

  18. Fluid-mixing studies in a hexagonal 37-pin wire-wrapped rod bundle

    International Nuclear Information System (INIS)

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

    1982-02-01

    Flow-split, pressure-drop, and mixing experiments were performed on a 37-pin LMFBR rod bundle with a P/D = 1.154 and H/D = 13.4 to verify the Chiu-Hawley-Burns correlations and to supplement the existing data base. The isokinetic extraction method, pitot-static probe pressure-measurement method, and salt-tracer-injection method were used for these experiments. The experimental results of the turbulent-flow-split parameters were predicted by the correlations within 3%. However, significant discrepancy between data and correlation existed in the transition flow regime (Re/sub b/ < 10,000). Flow-split parameters for Re/sub b/ < 3000 were not attainable because of the restriction of the isokinetic extraction method. The friction factor results showed a smooth transition from the laminar-flow regime to turbulent-flow regime. They were slightly overpredicted by the correlations, especially in the laminar-flow regime. The local swirl-flow ratio, C/sub IL/, in the turbulent-flow regime was found to be about 0.28, which was within 10% of the correlation value 0.265

  19. Assessment of CCFL model of RELAP5/MOD3 against simple vertical tubes and rod bundle tests

    International Nuclear Information System (INIS)

    Cho, Sung Jae; Arne, Nam Sung; Chung, Bub Dong; Kim, Hho Jung

    1991-01-01

    The CCFL model used in RELAP5/MOD3 version 5m5 has been assessed against simple vertical tubes and rod bundle tests performed at a facility of Korea Atomic Energy Research Institute. The effect of changes in tube diameter and nodalization of tube section were investigated. The roles of interfacial drags on the flooding characteristics are discussed. Difference between the calculation and the experiment are also discussed. A comparison between model assessment results and the test data showed that the calculated value lay well on the experimental flooding curve specified by user, but the pressure jump before onset of flooding was not calculated

  20. Heat transfer characteristics around a single heated rod immersed in sodium pool with gas jet injection

    International Nuclear Information System (INIS)

    Hideto Niikura; Kazuo Soga; Ken-ichiro Sugiyama; Akira Yamaguchi

    2005-01-01

    In a steam generator using liquid sodium, water intensely reacts with sodium when it leaks out from a heat transfer tube. It is important to evaluate the influence of sodium-water reaction to surrounding tubes and the shell. Hence, it has been desired to develop the simulation code for the evaluation of sodium-water reaction. From this viewpoint, the Japan Nuclear Cycle is now developing the SERAPHIM code. We reported a preliminary study to establish an experimental method for a single heated rod immersed in sodium pool with steam jet impingement planned in the near future as well as to obtain a preliminary data to verify the adequacy of SERAPHIM code. We first measured local and mean heat transfer coefficients around a horizontal single heated rod immersed in a water pool and a sodium pool with a limited volume in the experimental apparatus. It was confirmed that the mean heat transfer coefficients fairly agreed with the existing data for natural convection in water and sodium. Secondary we measured local and mean heat transfer coefficients around a horizontal single heated rod with Ar gas jet impingement immersed in the limited water pool and in the limited sodium pool. It was clearly observed that the local heat transfer coefficients in the sodium pool keep almost the same values in every angle regardless of increase in Ar gas jet velocity varied from about 8.7m/s to about 78m/s. On the other hand, it was confirmed in the water pool that local heat transfer coefficients on the forward stagnation side exposed in the Ar gas jet impingement increase with increasing the jet velocity while the local heat transfer coefficients on the opposite surface keep almost same values regardless of increase in the velocity. (authors)

  1. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    International Nuclear Information System (INIS)

    Cheng, Z.; Rao, Y.F.; Waddington, G.M.

    2014-01-01

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles

  2. IFPE/MT4-MT6A-LOCA, Large-break LOCA in-reactor fuel bundle materials tests at NRU

    International Nuclear Information System (INIS)

    Cunningham, Mitchel E.; Turnbull, J.A.

    2003-01-01

    generally presented in the reports on the tests. After the experiments, the test train was dismantled and cladding rupture sites were determined and fuel rod profilometry was performed in the spent fuel pool. Only limited destructive post-irradiation examination was performed on these two tests. Design and Objectives: - MT-4: The primary objectives of the MT-4 test included providing sufficient time in the alpha-Zircaloy ballooning window of 1033 to 1200 K to allow the 12 pressurized test rods to rupture before reflood cooling was introduced, obtaining data to determine heat transfer coefficients for ballooned and ruptured rods, and measuring rod internal gas pressure during rod deformation. All of the objectives for the test were accomplished. The MT-4 test bundle simulated a 6 x 6 section of a 17 x 17 PWR fuel assembly. There were 20 non-pressurized guard fuel rods to isolate the 12 central, pressurized tests rods; the four corner rods were deleted. The 12 test rods were fresh rods while the 20 guard rods had been used in a previous tests. Basic design information for the bundle and the 12 test rods is provided. - MT-6: A principal difference between MT-6A and the other tests was a redesign of the test train to reduce cladding circumferential temperature gradients and thus induce greater amounts of cladding ballooning and flow blockage. In addition, the 20 guard rods used in the previous tests were replaced with nine pressurized rods that had been used in a previous test. Thus, a total of 21 test rods were in MT-6A. Basic design information for the bundle and the test rods is provided. A malfunction of the computer controlling the test occurred during the test. As a result of this malfunction, system pressure during the transient heat-up was not at 0.28 MPa but was at 1.72 MPa. In addition, the desired temperature control was not achieved. This test was intended to provide the fuel cladding sufficient time in the a-Zircaloy temperature region (1050-1140 K) to maximize

  3. Effects of sleeve blockages on axial velocity and intensity of turbulence in an unheated 7 x 7 rod bundle

    International Nuclear Information System (INIS)

    Creer, J.M.; Rowe, D.S.; Bates, J.M.; Sutey, A.M.

    1976-01-01

    An experimental study is described which was performed to investigate the turbulent flow phenomena near postulated sleeve blockages in a model nuclear fuel rod bundle. The sleeve blockages were characteristic of fuel clad ''swelling'' or ''ballooning'' which could occur during loss-of-coolant accidents (LOCA) in pressurized water reactors. The study was conducted to provide information relative to the flow phenomena near postulated blockages to support detailed safety analyses of LOCAs. The results of the study are especially useful for verification of the hydraulic treatment of reactor core computer programs such as COBRA

  4. Hydrodynamics of vapor-liquid annular dispersed flows in channels with heated rod clusters under unsteady conditions

    International Nuclear Information System (INIS)

    Kroshilin, A.E.; Kroshilin, V.E.; Nigmatulin, B.I.

    1984-01-01

    A one-dimensional unsteady hydrodynamic model of vapour-liquid disperse-annular flows in channels with heated fuel rod clusters has been constructed. Regularities in the appearance of critical heat transfer due to the dryout of a near-wall liquid film on rod surfaces in such channels are investigated. The model developed takes into account the main flow regularities in the channels with heated rod clusters. The calculations made have shown that the time before crisis appearance agrees satisfactorily with the experimental data

  5. Assessment of fluid-to-fluid modelling of critical heat flux in horizontal 37-element bundle flows

    International Nuclear Information System (INIS)

    Yang, S.K.

    2006-01-01

    Fluid-to-fluid modelling laws of critical heat flux (CHF) available in the literature were reviewed. The applicability of the fluid-to-fluid modelling laws was assessed using available data ranging from low to high mass fluxes in horizontal 37-element bundles simulating a CANDU fuel string. Correlations consisting of dimensionless similarity groups were derived using modelling fluid data (Freon-12) to predict water CHF data in horizontal 37-element bundles with uniform and non-uniform axial-heat flux distribution (AFD). The results showed that at mass fluxes higher than ∼4,000 kg/m 2 s (water equivalent value), the vertical fluid-to-fluid modelling laws of Ahmad (1973) and Katto (1979) predict water CHF in horizontal 37-element bundles with non-uniform AFD with average errors of 1.4% and 3.0% and RMS errors of 5.9% and 6.1%, respectively. The Francois and Berthoud (2003) fluid-to-fluid modelling law predicts CHF in non-uniformly heated 37-element bundles in the horizontal orientation with an average error of 0.6% and an RMS error of 10.4% over the available range of 2,000 to 6,200 kg/m 2 s. (author)

  6. Design requirement on KALIMER control rod assembly duct

    International Nuclear Information System (INIS)

    Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J.

    1998-03-01

    This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs

  7. Design requirement on KALIMER control rod assembly duct

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, W.; Kang, H. Y.; Nam, C.; Kim, J. O.; Kim, Y. J

    1998-03-01

    This document establishes the design guidelines which are needs for designing the control rod assembly duct of the KALIMER as design requirements. it describes control rod assembly duct of the KALIMER and its requirements that includes functional requirements, performance requirements, interfacing systems, design limits and strength requirements, seismic requirements, structural requirements, environmental requirements, reliability and safety requirements, standard and codes, QA programs, and other requirements. The control rod system consists of three parts, which are drive mechanism, drive-line, and absorber bundle. This report deals with the absorber bundle and its outer duct only because the others are beyond the scope of fuel system design. The guidelines for design requirements intend to be used for an improved design of the control rod assembly duct of the KALIMER. (author). 19 refs.

  8. Hydraulic characteristics of HANARO fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S; Chung, H J; Chun, S Y; Yang, S K; Chung, M K [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    This paper presents the hydraulic characteristics measured by using LDV (Laser Doppler Velocimetry) in subchannels of HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regarding the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented. 9 refs., 12 figs. (Author)

  9. Hydraulic characteristics of HANARO fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cho, S.; Chung, H. J.; Chun, S. Y.; Yang, S. K.; Chung, M. K. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents the hydraulic characteristics measured by using LDV (Laser Doppler Velocimetry) in subchannels of HANARO, KAERI research reactor, fuel bundle. The fuel bundle consists of 18 axially finned rods with 3 spacer grids, which are arranged in cylindrical configuration. The effects of the spacer grids on the turbulent flow were investigated by the experimental results. Pressure drops for each component of the fuel bundle were measured, and the friction factors of fuel bundle and loss coefficients for the spacer grids were estimated from the measured pressure drops. Implications regarding the turbulent thermal mixing were discussed. Vibration test results measured by using laser vibrometer were presented. 9 refs., 12 figs. (Author)

  10. In-pool damaged fuel bundle recovery

    International Nuclear Information System (INIS)

    Piascik, T.G.; Patenaude, R.S.

    1988-01-01

    While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power / Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved

  11. COBRA - 3C/KFKI: a digital computer program for steady and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements

    International Nuclear Information System (INIS)

    Vigassy, J.; Kovacs, L.M.

    1977-11-01

    COBRA-3C/KFKI is a digital computer program for the CDC-3300 computer in FORTRAN language. The program is a revised version of the original COBRA-3C code. The code calculates steady-state and transient flow and enthalpy transport in rod-bundle nuclear fuel elements in both boiling and nonboiling conditions. The mathematical model is formulated by dividing the bundle flow area into flow subchannels that are assumed to contain one-dimensional flow and are coupled to each other by turbulent and diversion crossflow mixing. The program neglects sonic velocity propagation but allows for a temporal and spatial acceleration of the diversion crossflow in the transverse momentum equation. A semiexplicit finite-difference scheme is used to perform a boundary-value solution where the boundary conditions are the inlet enthalpy, inlet flow rate and exit pressure. (D.P.)

  12. Critical heat flux experiments for high conversion light water reactor, (3)

    International Nuclear Information System (INIS)

    Iwamura, Takamichi; Okubo, Tsutomu; Suemura, Takayuki; Hiraga, Fujio; Murao, Yoshio

    1990-03-01

    As a part of the thermal-hydraulic feasibility study of a high conversion light water reactor (HCLWR), critical heat flux (CHF) experiments were performed using triangular array rod bundles under steady-state and flow reduction transient conditions. The geometries of test sections were: rod outer diameter 9.5 mm, number of rods 4∼7, heated length 0.5∼1.0 m, and pitch to diameter ratio (P/D) 1.126∼1.2. The simulated fuel rod was a stainless steel tube and uniformly heated electrically with direct current. In the steady-state tests, pressures ranged: 1.0∼3.9 Mpa, mass velocities: 460∼4270 kg/s·m 2 , and exit qualities: 0.02∼0.35. In the transient tests, the times to CHF detection ranged from 0.5 to 25.4 s. The steady-state CHF's for the 4-rod test sections were higher than those for the 7-rod test sections with respect to the bundle averaged flow conditions. The measured CHF's increased with decreasing the heated length and decreased with decreasing the P/D. Based on the local flow conditions obtained with the subchannel analysis code COBRA-IV-I, KfK correlation agreed with the CHF data within 20 %, while WSC-2, EPRI-B and W, EPRI-Columbia and Kattor correlations failed to give satisfactory agreements. Under flow reduction rates less than 6 %/s, no significant difference in the onset conditions of DNB (departure from nucleate boiling) was recognized between the steady-state and transient conditions. At flow reduction rates higher than 6 %/s, on the other hand, the DNB occurred earlier than the DNB time predicted with the steady-state experiments. (author)

  13. Definition of the local fields of velocity, temperature and turbulent characteristics for axial stabilized fluid in arbitrary formed rod bundle assemblies

    International Nuclear Information System (INIS)

    Sedov, A.A.; Gagin, V.L.

    1995-01-01

    For the temperature fields in rod clads of experimental assemblies a good agreement have been got with use of prior calculations by subchannel code COBRA-IV-I, from results of which an additional information about δt/δX 3 distribution was taken. The method of definition the local fields of velocity, turbulent kinetic energy, temperature and eddy diffusivities for one-phase axial stabilized fluids in arbitrary formed rod bundle assemblies with invariable upward geometry was developed. According to this model the AGURA code was worked out to calculate local thermal hydraulic problems in combination with temperature fields in fuel rods and constructive elements of fuel assemblies. The method does not use any prior geometric scales and is based only on invariant local flow parameters: turbulent kinetic energy, velocity field deformation tensor and specific work of inner friction. Verification of this method by available experimental data showed a good agreement of calculation data and findings of velocity and t.k.e. fields, when the secondary flows have not a substantial influence to a balance of axial momentum and turbulent kinetic energy. (author)

  14. Verification and validation of a numeric procedure for flow simulation of a 2x2 PWR rod bundle

    International Nuclear Information System (INIS)

    Santos, Andre A.C.; Barros Filho, Jose Afonso; Navarro, Moyses A.

    2011-01-01

    Before Computational Fluid Dynamics (CFD) can be considered as a reliable tool for the analysis of flow through rod bundles there is a need to establish the credibility of the numerical results. Procedures must be defined to evaluate the error and uncertainty due to aspects such as mesh refinement, turbulence model, wall treatment and appropriate definition of boundary conditions. These procedures are referred to as Verification and Validation (V and V) processes. In 2009 a standard was published by the American Society of Mechanical Engineers (ASME) establishing detailed procedures for V and V of CFD simulations. This paper presents a V and V evaluation of a numerical methodology applied to the simulation of a PWR rod bundle segment with a split vane spacer grid based on ASMEs standard. In this study six progressively refined meshes were generated to evaluate the numerical uncertainty through the verification procedure. Experimental and analytical results available in the literature were used in this study for validation purpose. The results show that the ASME verification procedure can give highly variable predictions of uncertainty depending on the mesh triplet used for the evaluation. However, the procedure can give good insight towards optimization of the mesh size and overall result quality. Although the experimental results used for the validation were not ideal, through the validation procedure the deficiencies and strengths of the presented modeling could be detected and reasonably evaluated. Even though it is difficult to obtain reliable estimates of the uncertainty of flow quantities in the turbulent flow, this study shows that the V and V process is a necessary step in a CFD analysis of a spacer grid design. (author)

  15. Heat flux and temperature determination on the control rod outer surface

    International Nuclear Information System (INIS)

    Taler, J.; Cebula, A.; Marcinkiewicz, J.; Tinoco, H.

    2011-01-01

    The paper presents heat transfer calculation results concerning a control rod of Unit 3 of Forsmark Nuclear Power Plant (NPP). The part of the control rod, which is the object of interest, is surrounded by a mixing region of hot and cold flows and, as a consequence, is subjected to thermal fluctuations. The paper describes a numerical test which validates the method based on the solution of the inverse heat conduction problem (IHCP). The comparison of the results achieved by two methods, CFD and IHCP, including a description of the IHCP method used in the calculation process, shows a very good agreement between the methods. (author)

  16. CFD modeling of turbulent mixing through vertical pressure tube type boiling water reactor fuel rod bundles with spacer-grids

    Science.gov (United States)

    Verma, Shashi Kant; Sinha, S. L.; Chandraker, D. K.

    2018-05-01

    Numerical simulation has been carried out for the study of natural mixing of a Tracer (Passive scalar) to describe the development of turbulent diffusion in an injected sub-channel and, afterwards on, cross-mixing between adjacent sub-channels. In this investigation, post benchmark evaluation of the inter-subchannel mixing was initiated to test the ability of state-of-the-art Computational Fluid Dynamics (CFD) codes to numerically predict the important turbulence parameters downstream of a ring type spacer grid in a rod-bundle. A three-dimensional Computational Fluid Dynamics (CFD) tool (STAR-CCM+) was used to model the single phase flow through a 30° segment or 1/12th of the cross segment of a 54-rod bundle with a ring shaped spacer grid. Polyhedrons were used to discretize the computational domain, along with prismatic cells near the walls, with an overall mesh count of 5.2 M cell volumes. The Reynolds Stress Models (RSM) was tested because of RSM accounts for the turbulence anisotropy, to assess their capability in predicting the velocities as well as mass fraction of potassium nitrate measured in the experiment. In this way, the line probes are located in the different position of subchannels which could be used to characterize the progress of the mixing along the flow direction, and the degree of cross-mixing assessed using the quantity of tracer arriving in the neighbouring sub-channels. The predicted dimensionless mixing scalar along the length, however, was in good agreement with the measurements downstream of spacers.

  17. Synthesis of the turbulent mixing in a rod bundle with vaned spacer grids based on the OECD-KAERI CFD benchmark exercise

    International Nuclear Information System (INIS)

    Lee, Jae Ryong; Kim, Jungwoo; Song, Chul-Hwa

    2014-01-01

    Highlights: • OECD/KAERI international CFD benchmark exercise was operated by KAERI. • The purpose is to validate relevant CFD codes based on the MATiS-H experiments. • Blind calculation results were synthesized in terms of mean velocity and RMS. • Quality of control volume rather than the number of it was emphasized. • Major findings were followed OECD/NEA CSNI report. - Abstract: The second international CFD benchmark exercise on turbulent mixing in a rod bundle has been launched by OECD/NEA, to validate relevant CFD (Computational Fluid Dynamics) codes and develop problem-specific Best Practice Guidelines (BPG) based on the KAERI (Korea Atomic Energy Research Institute) MATiS-H experiments on the turbulent mixing in a 5 × 5 rod array having two different types of vaned spacer grids: split and swirl types. For this 2nd international benchmark exercise (IBE-2), the MATiS-H testing provided a unique set of experimental data such as axial and lateral velocity components, turbulent intensity, and vorticity information. Blind CFD calculation results were submitted by twenty-five (25) participants to KAERI, who is the host organization of the IBE-2, and then analyzed and synthesized by comparing them with the MATiS-H data. Based on the synthesis of the results from both the experiments and blind CFD calculations for the IBE-2, and also by comparing with the IBE-1 benchmark exercise on the mixing in a T-junction, useful information for simulating this kind of complicated physical problem in a rod bundle was obtained. And some additional Best Practice Guidelines (BPG) are newly proposed. A summary of the synthesis results obtained in the IBE-2 is presented in this paper

  18. Synthesis of the turbulent mixing in a rod bundle with vaned spacer grids based on the OECD-KAERI CFD benchmark exercise

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Ryong; Kim, Jungwoo; Song, Chul-Hwa, E-mail: chsong@kaeri.re.kr

    2014-11-15

    Highlights: • OECD/KAERI international CFD benchmark exercise was operated by KAERI. • The purpose is to validate relevant CFD codes based on the MATiS-H experiments. • Blind calculation results were synthesized in terms of mean velocity and RMS. • Quality of control volume rather than the number of it was emphasized. • Major findings were followed OECD/NEA CSNI report. - Abstract: The second international CFD benchmark exercise on turbulent mixing in a rod bundle has been launched by OECD/NEA, to validate relevant CFD (Computational Fluid Dynamics) codes and develop problem-specific Best Practice Guidelines (BPG) based on the KAERI (Korea Atomic Energy Research Institute) MATiS-H experiments on the turbulent mixing in a 5 × 5 rod array having two different types of vaned spacer grids: split and swirl types. For this 2nd international benchmark exercise (IBE-2), the MATiS-H testing provided a unique set of experimental data such as axial and lateral velocity components, turbulent intensity, and vorticity information. Blind CFD calculation results were submitted by twenty-five (25) participants to KAERI, who is the host organization of the IBE-2, and then analyzed and synthesized by comparing them with the MATiS-H data. Based on the synthesis of the results from both the experiments and blind CFD calculations for the IBE-2, and also by comparing with the IBE-1 benchmark exercise on the mixing in a T-junction, useful information for simulating this kind of complicated physical problem in a rod bundle was obtained. And some additional Best Practice Guidelines (BPG) are newly proposed. A summary of the synthesis results obtained in the IBE-2 is presented in this paper.

  19. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    International Nuclear Information System (INIS)

    Rao, Y.F.; Cheng, Z.; Waddington, G.M.

    2014-01-01

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles

  20. Natural convection heat transfer for a staggered array of heated, horizontal cylinders within a rectangular enclosure

    Energy Technology Data Exchange (ETDEWEB)

    Triplett, C.E.

    1996-12-01

    This thesis presents the results of an experimental investigation of natural convection heat transfer in a staggered array of heated cylinders, oriented horizontally within a rectangular enclosure. The main purpose of this research was to extend the knowledge of heat transfer within enclosed bundles of spent nuclear fuel rods sealed within a shipping or storage container. This research extends Canaan`s investigation of an aligned array of heated cylinders that thermally simulated a boiling water reactor (BWR) spent fuel assembly sealed within a shipping or storage cask. The results are presented in terms of piecewise Nusselt-Rayleigh number correlations of the form Nu = C(Ra){sup n}, where C and n are constants. Correlations are presented both for individual rods within the array and for the array as a whole. The correlations are based only on the convective component of the heat transfer. The radiative component was calculated with a finite-element code that used measured surface temperatures, rod array geometry, and measured surface emissivities as inputs. The correlation results are compared to Canaan`s aligned array results and to other studies of natural convection in horizontal tube arrays.

  1. Natural convection heat transfer for a staggered array of heated, horizontal cylinders within a rectangular enclosure

    International Nuclear Information System (INIS)

    Triplett, C.E.

    1996-12-01

    This thesis presents the results of an experimental investigation of natural convection heat transfer in a staggered array of heated cylinders, oriented horizontally within a rectangular enclosure. The main purpose of this research was to extend the knowledge of heat transfer within enclosed bundles of spent nuclear fuel rods sealed within a shipping or storage container. This research extends Canaan's investigation of an aligned array of heated cylinders that thermally simulated a boiling water reactor (BWR) spent fuel assembly sealed within a shipping or storage cask. The results are presented in terms of piecewise Nusselt-Rayleigh number correlations of the form Nu = C(Ra) n , where C and n are constants. Correlations are presented both for individual rods within the array and for the array as a whole. The correlations are based only on the convective component of the heat transfer. The radiative component was calculated with a finite-element code that used measured surface temperatures, rod array geometry, and measured surface emissivities as inputs. The correlation results are compared to Canaan's aligned array results and to other studies of natural convection in horizontal tube arrays

  2. The Vibration Analysis of Tube Bundles Induced by Fluid Elastic Excitation in Shell Side of Heat Exchanger

    Science.gov (United States)

    Bao, Minle; Wang, Lu; Li, Wenyao; Gao, Tianze

    2017-09-01

    Fluid elastic excitation in shell side of heat exchanger was deduced theoretically in this paper. Model foundation was completed by using Pro / Engineer software. The finite element model was constructed and imported into the FLUENT module. The flow field simulation adopted the dynamic mesh model, RNG k-ε model and no-slip boundary conditions. Analysing different positions vibration of tube bundles by selecting three regions in shell side of heat exchanger. The results show that heat exchanger tube bundles at the inlet of the shell side are more likely to be failure due to fluid induced vibration.

  3. Thermal hydraulic calculation of wire-wrapped bundles using a finite element method. Thesee code

    International Nuclear Information System (INIS)

    Rouzaud, P.; Gay, B.; Verviest, R.

    1981-07-01

    The physical and mathematical models used in the THESEE code now under development by the CEA/CEN Cadarache are presented. The objective of this code is to predict the fine three-dimensional temperature field in the sodium in a wire-wrapped rod bundle. Numerical results of THESEE are compared with measurements obtained by Belgonucleaire in 1976 in a sodium-cooled seven-rod bundle

  4. Assembly mechanism for nuclear fuel bundles

    International Nuclear Information System (INIS)

    Long, J.W.; Flora, B.S.

    1977-01-01

    A method of securing a fuel bundle to permit easy remote disassembly is described. Fuel rods are held loosely between end plates, each end of the rods fitting into holes in the end plates. At the upper end of each fuel rod there is a spring pressing against the end plate. Tie rods are used to hold the end plates together securely. The lower end of each tie rod is screwed into the lower end plate; the upper end of each tie rod is attached to the upper end plate by means of a locking assembly described in the patent. In order to remove the upper tie plate during the disassembly process, it is necessary only to depress the tie plate against the pressure of the springs surrounding the fuel rods and then to rotate each locking sleeve on the tie rods from its locked to its unlocked position. It is then possible to remove the tie plate without disassembling the locking assembly. (LL)

  5. LMFR core and heat exchanger thermohydraulic design: former USSR and present Russian approaches

    International Nuclear Information System (INIS)

    1999-01-01

    The information presented in this report is dealing with liquid metal cooled fast reactors some of which are in operation (France, Japan, Russian federation) or under construction. Comprehensive thermal hydraulic research both experimental and numeric applied to such reactors was carried out in the Institute of Physics and Power Engineering (IPPE), Obninsk, Russian Federation. The IAEA Working Group on fast Reactors (IWGFR) recommended that IPPE should generalize its thermal hydraulic studies as well as results of other countries published previously in the field of liquid metal flow distribution and heat transfer in fuel pin and heat exchanger rod bundles (France, Germany, Japan, India, Russian Federation, United Kingdom and United States). The validity of computer codes and design approaches was proven by comparison of calculated results with measured values of velocity, pressure, temperature distributions in rod bundles cooled/heated by liquid metal, usually sodium. The report includes the methodology and philosophy of the analytical and experimental investigations when applied to core and heat exchanger thermal hydraulic design of Light Water Moderated Fast Reactors (LMFRs)

  6. Fuel rod simulator effects in flooding experiments single rod tests

    International Nuclear Information System (INIS)

    Nishida, M.

    1984-09-01

    The influence of a gas filled gap between cladding and pellet on the quenching behavior of a PWR fuel rod during the reflood phase of a LOCA has been investigated. Flooding experiments were conducted with a short length electrically heated single fuel rod simulator surrounded by glass housing. The gap of 0.05 mm width between the Zircaloy cladding and the internal Al 2 O 3 pellets of the rod was filled either wit helium or with argon to vary the radial heat resistance across the gap. This report presents some typical data and an evaluation of the reflood behavior of the fuel rod simulator used. The results show that the quench front propagates faster for increasing heat resistance in the gap between cladding and heat source of the rod. (orig.) [de

  7. Multi-rod burst behavior under a loss-of-coolant accident condition, (1)

    International Nuclear Information System (INIS)

    Hashimoto, Masao; Otomo, Takashi; Furuta, Teruo; Kawasaki, Satoru; Uetsuka, Hiroshi

    1980-12-01

    Multi-rod burst tests have been planned since 1977 to estimate quantitative channel restriction during a LOCA transient in LWRs. For this purpose, many bundle tests have been making to burst in a steam in varying a few parameters which influence the degree of channel restriction. The purpose of this report is to provide a background document for final reports to be published in the future. This report includes the results of No. 7805 bundle test, namely temperature, internal pressure, burst behavior of rods and channel restriction of the bundle. (author)

  8. Microstructural engineering applied to the controlled cooling of steel wire rod: Part I. Experimental design and heat transfer

    Science.gov (United States)

    Campbell, P. C.; Hawbolt, E. B.; Brimacombe, J. K.

    1991-11-01

    The goal of this study was to develop a mathematical model which incorporates heat flow, phase transformation kinetics, and property-structure-composition relationships to predict the mechanical properties of steel rod being control cooled under industrial conditions. Thus, the principles of microstructural engineering have been brought to bear on this interdisciplinary problem by combining computer modeling with laboratory measurements of heat flow, austenite decomposition kinetics, microstructure and mechanical properties, and industrial trials to determine heat transfer and obtain rod samples under known conditions. Owing to the length and diversity of the study, it is reported in three parts,[8191]the first of which is concerned with the heat flow measurements. A relatively simple and reliable technique, involving a preheated steel rod instrumented with a thermocouple secured at its centerline, has been devised to determine the cooling rate in different regions of the moving bed of rod loops on an operating Stelmor line. The measured thermal response of the rod has been analyzed by two transient conduction models (lumped and distributed parameter, respectively) to yield overall heat-transfer coefficients for radiation and convection. The adequacy of the technique has been checked by cooling instrumented rods under well-defined, air crossflow conditions in the laboratory and comparing measured heat-transfer coefficients to values predicted from well-established equations. The industrial thermal measurements have permitted the characterization of a coefficient to account for radiative interaction among adjacent rod loops near the edge and at the center of the bed.

  9. Heat transfer to a dispersed two-phase flow and detailed quench front velocity research

    International Nuclear Information System (INIS)

    Boer, T.C. de; Molen, S.B. van der

    1985-01-01

    A programme to obtain a data base for 'Boildown and Reflood' computer code development and to obtain information on the influence of non-uniform temperature and/or power profile on the quench front velocity and prequench heat transfer, including unheated wall and grid effects, has been undertaken. It is in two parts. In the first (for the tube, annulus and a 4-rod bundle) an early wetting of the unheated shroud is shown. This leads to an increase in quench front velocity and in liquid transport downstream from the quench front. For the inverted annular flow regime the extended Bromley correlation gives good agreement with the experimental data. In the second part (36-rod bundle reflood test programme) the wall-temperature differences in the radial direction gives rise to heat transfer processes which are described and explained. (U.K.)

  10. Experimental investigation of water sprayed finned heat exchanger tube bundles

    International Nuclear Information System (INIS)

    Sommer, A.

    1987-07-01

    Experimental investigations have been made to study the performance of two finned tube-bundle heat exchangers (FORGO type) when wetted by water sprays. The heat exchangers are designed to cool water in a dry cooling tower. The test-elements had a frontal area of 1 m 2 . The water sprays were created by 20 nozzles, 200 mm in front of the heat exchangers. Air velocities at the inlet of the coolers were in the range 0,8 m/s to 12 m/s and initial temperature differences ITD reached 45 degrees C. The test facility was designed to determine the combined latent and sensible heat fluxes in the wetted heat exchanger, the airside pressure drop and the air humidity and temperature at the exchanger inlet and outlet, and to measure the weight of the water wetting the cooler's surface. The sprayed test elements were investigated in different positions, but most of the experiments were carried out in the position with the fins horizontal

  11. Fuel bundle for nuclear reactor

    International Nuclear Information System (INIS)

    Long, J.W.; Flora, B.S.; Ford, K.L.

    1977-01-01

    The invention concerns a new, simple and inexpensive system for assembling and dismantling a nuclear reactor fuel bundle. Several fuel rods are fitted in parallel rows between two retaining plates which secure the fuel rods in position and which are maintained in an assembled position by means of several stays fixed to the two end plates. The invention particularly refers to an improved apparatus for fixing the stays to the upper plate by using locking fittings secured to rotating sleeves which are applied against this plate [fr

  12. Single-Phase Bundle Flows Including Macroscopic Turbulence Model

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Jun; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of); Yoon, Seok Jong; Cho, Hyoung Kyu [Seoul National University, Seoul (Korea, Republic of)

    2016-05-15

    To deal with various thermal hydraulic phenomena due to rapid change of fluid properties when an accident happens, securing mechanistic approaches as much as possible may reduce the uncertainty arising from improper applications of the experimental models. In this study, the turbulence mixing model, which is well defined in the subchannel analysis code such as VIPRE, COBRA, and MATRA by experiments, is replaced by a macroscopic k-e turbulence model, which represents the aspect of mathematical derivation. The performance of CUPID with macroscopic turbulence model is validated against several bundle experiments: CNEN 4x4 and PNL 7x7 rod bundle tests. In this study, the macroscopic k-e model has been validated for the application to subchannel analysis. It has been implemented in the CUPID code and validated against CNEN 4x4 and PNL 7x7 rod bundle tests. The results showed that the macroscopic k-e turbulence model can estimate the experiments properly.

  13. Bundle duct interaction studies for fuel assemblies

    International Nuclear Information System (INIS)

    Hsia, H.T.S.; Kaplan, S.

    1981-06-01

    It is known that the wire-wrapped rods and duct in an LMFBR are undergoing a gradual structural distortion from the initially uniform geometry under the combined effects of thermal expansion and irradiation induced swelling and creep. These deformations have a significant effect on flow characteristics, thus causing changes in thermal behavior such as cladding temperature and temperature distribution within a bundle. The temperature distribution may further enhance or retard irradiation induced deformation of the bundle. This report summarizes the results of the continuing effort in investigating the bundle-duct interaction, focusing on the need for the large development plant

  14. SEU43 fuel bundles in CANDU 600

    International Nuclear Information System (INIS)

    Catana, Alexandru; Prodea, Iosif; Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

    2008-01-01

    Cernavoda Unit 1 and Unit 2 are pressure tube 650 MWe nuclear stations moderated and cooled with heavy water, of Canada design, located in Romania. Fuelling is on-power and the plant is currently fuelled with natural uranium dioxide. Fuel is encapsulated in a 37 fuel rod assembly having a specific standard geometry (STD37). In order to reduce fuel cycle costs programs were initiated in Canada, South Korea and at SCN Pitesti, Romania for design and build of a new, improved geometry fuel bundle and some fuel compositions. Among fuel compositions, which are considered, is the slightly enriched uranium (SEU) fuel (0.96 w% U-235) with an associated burn-up increase from ∼7900 MWd/tU up to ∼15000 MWd/tU. Neutron analysis showed that the Canadian-Korean fuel bundle geometry with 43 rods called SEU (SEU43) can be used in already operated reactors. A new fuel bundle resulted. Extended, comprehensive analysis must be conducted in order to assess the TH behavior of SEU43 besides the neutron, mechanical (drag force, etc) analyses. In this paper, using the sub-channel approach, main thermal-hydraulic parameters were analyzed: pressure drop; fuel, sheath and coolant temperatures; coolant density; critical heat flux. Some significant differences versus standard fuel are outlined in the paper and some conclusions are drawn. While, by using this new fuel, there are many benefits to be attained like: fuel costs reduction, spent fuel waste minimization, increase in competitiveness of nuclear power generation against other sources of generation, etc., the safety margins must be, at least, conserved. The introduction of a new fuel bundle type, different in geometry and fuel composition, requires a detailed preparation, a testing program and a series of neutron and thermal-hydraulic analysis. The results reported by this paper is part of this effort. The feasibility to increase the enrichment from 0.71% U-235 (NU) to 0.96% U-235, with an estimated burn-up increase up to 14000 MWd

  15. ASSERT and COBRA predictions of flow distribution in vertical bundles

    International Nuclear Information System (INIS)

    Tahir, A.; Carver, M.B.

    1983-01-01

    COBRA and ASSERT are subchannel codes which compute flow and enthalpy distributions in rod bundles. COBRA is a well known code, ASSERT is under development at CRNL. This paper gives a comparison of the two codes with boiling experiments in vertical seven rod bundles. ASSERT predictions of the void distribution are shown to be in good agreement with reported experimental results, while COBRA predictions are unsatisfactory. The mixing models in both COBRA and ASSERT are briefly discussed. The reasons for the failure of COBRA-IV and the success of ASSERT in simulating the experiments are highlighted

  16. Simulation of the fuel rod bundle test QUENCH-03 using the system codes ASTEC and ATHLET-CD

    International Nuclear Information System (INIS)

    Kruse, P.; Koch, M.K.

    2011-01-01

    The QUENCH-03 test was performed on the 21. of January 1999 at FZK (Forschungszentrum Karlsruhe) to investigate the behaviour on reflood of PWR (Pressurized Water Reactor) fuel rods with little oxidation. This paper presents the results of the simulation of QUENCH-03 performed with the version V1.3 of the integral code ASTEC (Accident Source Term Evaluation Code) which is being developed by IRSN (France) in cooperation with GRS (Germany) and with the program version 2.1A of the mechanistic code ATHLET-CD (Analysis of Thermal-hydraulics of Leaks and Transients - Core Degradation) which is under development by GRS. At first the QUENCH test facility and the QUENCH test program in general are described. The test conduct of the test QUENCH-03 follows as well as a description of the used codes ASTEC and ATHLET-CD with the associated modeling of the test section. The results of this calculation show that during the heat-up and transient phase both codes can calculate bundle and shroud temperatures as well as the hydrogen production in good approximation to the experimental data. During the quench phase and up to the end of the test only the oxidation model PRATER of ASTEC simulates the hydrogen production very well, the other oxidation models of ASTEC cannot calculate to some extent the measured amount of hydrogen. ATHLET-CD underestimates the integral amount at the end of the test. In the ASTEC calculations the temperatures during the quench phase show qualitatively good results, only time delays on some elevations of the bundle could be noticed. ATHLET-CD reproduces the thermal behaviour up to the first temperature escalation very well, after that the temperatures are partly over-estimated. The time delay recognized in the ASTEC calculations are seen as well. The results of the integral code ASTEC emphasize that the calculation of QUENCH-03 is possible and leading to good results concerning hydrogen release and corresponding temperatures. Because the QUENCH-03 test was

  17. Calculation of heat transfer in transversely stream-lined tube bundles with chess arrangement

    International Nuclear Information System (INIS)

    Migaj, V.K.

    1978-01-01

    A semiempirical theory of heat transfer in transversely stream-lined chess-board tube bundles has been developed. The theory is based on a single cylinder model and involves external flow parameter evaluation on the basis of the solidification principle of a vortex zone. The effect of turbulence is estimated according to experimental results. The method is extended to both average and local heat transfer coefficients. Comparison with experiment shows satisfactory agreement

  18. Application of field synergy principle for optimization fluid flow and convective heat transfer in a tube bundle of a pre-heater

    International Nuclear Information System (INIS)

    Hamid, Mohammed O.A.; Zhang, Bo; Yang, Luopeng

    2014-01-01

    The big problems facing solar-assisted MED (multiple-effect distillation) desalination unit are the low efficiency and bulky heat exchangers, which worsen its systematic economic feasibility. In an attempt to develop heat transfer technologies with high energy efficiency, a mathematical study is established, and optimization analysis using FSP (field synergy principle) is proposed to support meaning of heat transfer enhancement of a pre-heater in a solar-assisted MED desalination unit. Numerical simulations are performed on fluid flow and heat transfer characteristics in a circular and elliptical tube bundle. The numerical results are analyzed using the concept of synergy angle and synergy number as an indication of synergy between velocity vector and temperature gradient fields. Heat transfer in elliptical tube bundle is enhanced significantly with increasing initial velocity of the feed seawater and field synergy number and decreasing of synergy angle. Under the same operating conditions of the two designs, the total average synergy angle is 78.97° and 66.31° in circular and elliptical tube bundle, respectively. Optimization of the pre-heater by FSP shows that in case of elliptical tube bundle design, the average synergy number and heat transfer rate are increased by 22.68% and 35.98% respectively. - Highlights: • FSP (field synergy principle) is used to investigate heat transfer enhancement. • Numerical simulations are performed in circular and elliptical tubes pre-heater. • Numerical results are analyzed using concept of synergy angle and synergy number. • Optimization of elliptical tube bundle by FSP has better performance

  19. Suggestions for the interpretation of temperature noise measurements in a heated linear bundle in a water loop

    International Nuclear Information System (INIS)

    Schwalm, D.

    1975-09-01

    A concept is described how to use temperature noise for the detection and identification of a simulated malfunction (e.g. a blockage) in a heated linear bundle in the preboiling state. At first, methods are proposed how to find an optimal detector position down stream from the bundle exit in such a way that the detector sees the total bundle cross section. In addition some methods are proposed for the identification of the malfunction by making use of random data analysis

  20. DANCOFF-MC: A program to calculate Dancoff factors in CANDU type fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Feher, S; Valko, J

    1992-12-01

    The objective of DANCOFF-MC is the evaluation of Dancoff factors for cylindrical fuel rods arranged parallel in various and complicated bundle geometries. No interaction with fuel rods in any of the other bundles are considered due to the large distance, in mean free paths, between the buldes. Using a common basic algorithm three versions of the program have been written so far: The DANCOFF-MC-2, the DANCOFF-MC-19 and the DANCOFF-MC-27. (orig./HP).

  1. The safety feature of hydraulic driving system of control rod for 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Chi Zongbo; Wu Yuanqiang

    1997-01-01

    The hydraulic driving system of control rod is used as control rod drive mechanism in 200 MW nuclear heating reactor. Design of this system is based on passive system, integrating drive and guide of control rod. The author analyzes the inherent safety and the design safety of this system, with mechanism of control rod not ejecting when the pressure of pressure vessel is lost, and calculating result of core not exposing when the amount of coolant is drained by broken pipe. The results indicate that this system has good safety feature, and assures reactor safety under any accident conditions, providing important technology support for 200 MW nuclear heating reactor with inherent safety feature

  2. Laminar simulation of intersubchannel mixing in a triangular nuclear fuel bundle geometry

    International Nuclear Information System (INIS)

    Zaretsky, A.; Lightstone, M.F.; Tullis, S.

    2015-01-01

    Highlights: • Quasi-periodic flow was observed through rod-to-wall gaps. • Triangular subchannel flows were fundamentally irregular. • Cross-gap flow was influenced both by local and adjacent cross-gap intensity. • Phase-linking between gaps induced cross-plane peripheral circulation through rod–wall gaps. • Cross-gap flow structure was dependent on subchannel geometry. - Abstract: Predicting temperature distributions in fuel rod bundles is an important component of nuclear reactor safety analysis. Intersubchannel mixing acts to homogenize coolant temperatures thus reducing the likelihood of localized regions of high fuel temperature. Previous research has shown that intersubchannel mixing in nuclear fuel rod bundles is enhanced by a large-scale quasi-periodic energetic fluid motion, which transports fluid on the cross-plane between the narrow gaps connecting subchannels. This phenomenon has also been observed in laminar flows. Unsteady laminar flow simulations were performed in a simplified bundle of three rods with a pipe. Three similar geometries of varying gap width were examined, and a thermal trace was implemented on the first geometry. Thermal mixing was driven by the advection of energy between subchannels by the cross-plane flow. Flow through the rod-to-wall gaps in the wall subchannels alternated with a dominant frequency, particularly when rod-to-wall gaps were smaller than rod-to-rod gaps. Significant phase-linking between rod-to-wall gaps was also observed such that a peripheral circulation occurred through each gap simultaneously. Cross-plane flow through the rod-to-rod gaps in the triangular subchannel was irregular in each case. This was due to the fundamental irregularity of the triangular subchannel geometry. Vortices were continually broken up by cross-plane flow from other gaps due to the odd number of fluid pathways within the central subchannel. Cross-plane flow in subchannel geometries is highly interconnected between gaps. The

  3. Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data

    International Nuclear Information System (INIS)

    Panka, I.; Kereszturi, A.

    2014-01-01

    The assessment of the uncertainties of COBRA-IIIC thermal-hydraulic analyses of rod bundles is performed for a 5-by-5 bundle representing a PWR fuel assembly. In the first part of the paper the modeling uncertainties are evaluated in the term of the uncertainty of the turbulent mixing factor using the OECD NEA/NRC PSBT benchmark data. After that the uncertainties of the COBRA calculations are discussed performing Monte-Carlo type statistical analyses taking into account the modeling uncertainties and other uncertainties prescribed in the OECD NEA UAM benchmark specification. Both steady-state and transient cases are investigated. The target quantities are the uncertainties of the void distribution, the moderator density, the moderator temperature and the DNBR. We will see that - beyond the uncertainties of the geometry and the boundary conditions - it is very important to take into account the modeling uncertainties in case of bundle or sub-channel thermo-hydraulic calculations.

  4. Development and application of an asymmetric deformation model to describe the fuel rod behaviour during LOCA

    International Nuclear Information System (INIS)

    Chakraborty, A.K.; Schubert, J.D.

    1983-01-01

    For calculation of clad ballooning from single rod and rod bundle experiments a model considering the influences of azimuthal temperature gradients due to the existing eccentricity of the pellets has been developed. This model is based on the secondary creep model of Norton and on the concentric deformation model ending in cladding burst as proposed by F. Erbacher. The new model considers the azimuthal temperature differences along the cladding and the resulting differences in deformations. With this model, calculations of cladding burst deformations from single rod and rod bundle experiments are performed with good agreement

  5. Calculation Of A Lattice Physics Parameter For SBWR Fuel Bundle Design

    International Nuclear Information System (INIS)

    Sardjono, Y.

    1996-01-01

    The maximum power peaking factor for Nuclear Power Plant SBWR type is 1.5. The precision for that calculation is related with the result of unit cell analysis each rod in the fuel bundles. This analysis consist of lattice eigenvalue, lattice average diffusion cross section as well as relative power peaking factor in the fuel rod for each fuel bundles. The calculation by using TGBLA computer code which is based on the transport and 168 group diffusion theory. From this calculation can be concluded that the maximum relative power peaking factor is 1.304 and lower than design limit

  6. Analytical model for calculation of the thermo hydraulic parameters in a fuel rod assembly

    Energy Technology Data Exchange (ETDEWEB)

    Cesna, B., E-mail: benas@mail.lei.l [Lithuanian Energy Institute, Laboratory of Nuclear Installation Safety, Breslaujos g. 3, LT-44403 Kaunas (Lithuania)

    2010-11-15

    Research highlights: {yields} Proposed calculation model can be used for rapid calculation of the bundles with rods spaced by wire wrapping or honey type spacer grids. {yields} Model estimate three flow cross mixture mechanisms. {yields} Program DARS is enable to analyses experimental results. - Abstract: The paper presents the procedure of the cellular calculation of thermo hydraulic parameters of a single-phase gas flow in a fuel rod assembly. The procedure is implemented in the DARS program. The program is intended for calculation of the distribution of the gaseous coolant parameters and wall temperatures in case of arbitrary, geometrically specified, arrangement of the rods in fuel assembly and in case of arbitrary, functionally specified in space, heat release in the rods. In mathematical model the flow cross-section of the channel of intricate shape is conventionally divided to elementary cells formed by straight lines, which connect the centers of rods. Within the limits of a single cell the coolant parameters and the temperature of the corresponding part of the rod surface are assumed constant. The entire fuel assembly is viewed as a system of parallel interconnected channels. Program DARS is illustrated by calculation of a temperature mode of 85-rod assembly with spacers of wire wrapping on the rods.

  7. A critical heat flux correlation for advanced pressurized light water reactor application

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Hame, W.

    1982-05-01

    Many CHF-correlations have been developed for water cooled rod clusters representing typical PWR or BWR fuel element geometries with relative wide rod lattices. However the fuel elements of an Advanced Pressurized Water Reactor (APWR) have a tight fuel rod lattice, in view of increasing the fuel utilization. It was therefore decided to produce a new CHF-correlation valid for rod bundles with tight lattices. The already available WSC-2 correlation was chosen as a basis. The geometry dependent parameters of this correlation were determined again with the method of the root mean square fitting from the experimental data of the CHF-tests performed in the frame of the Light Water Breeder Reactor programme at the Bettis Laboratory. These tests include triangular array rod bundles with very tight lattices. Furthermore the effect of spiral spacer ribs was investigated on the basis of experimental data from the Columbia University. Application of the new CHF-correlation to conditions typical for an APWR shows that the predicted critical heat fluxes are much smaller than those calculated with the usual PWR-CHF-correlations, but they are higher than those predicted by the B+W-VPI+SU correlation. (orig.) [de

  8. Construction of supporting grids for fuel rods (or tubes in a heat exchanger)

    International Nuclear Information System (INIS)

    1975-01-01

    The construction of supporting grids for fuel rods (or tubes in heat exchangers) is described. It is a modification of a former French patent. The modification consists in the use of different material for the springs keeping the rod in place and describes another way of fixing these blade-shaped springs. Advantages of the specific spring characteristics were taken into consideration

  9. COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

    Energy Technology Data Exchange (ETDEWEB)

    Wheeler, C.L.; Stewart, C.W.; Cena, R.J.; Rowe, D.S.; Sutey, A.M.

    1976-03-01

    The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds of subchannels.

  10. Turbulent flow heat transfer in ET-RR-1

    International Nuclear Information System (INIS)

    Khattab, M.; Mina, A.R.

    1990-01-01

    In nuclear reactors the effect of heat transfer coefficient, which depends on the constant C. Is primordial in calculating the clad surface temperatures. To determine the constant C of ET-RR-1 fuel bundles based on in-pile measurements different well known and recommended values of C are verified. A computer program is written to calculate steady thermal core characteristics at different operating conditions. The total flow rate is distributed considering same pressure drop across the core irrespective of bundle location. The total reactor power is readily distributed as Bessel function. The flow and power per bundle are equally distributed among the fuel rods irrespective of their positions inside the bundle. It is found that the constant C equals 0.047 gives acceptable compatibility between measurements and calculations. The maximum clad surface temperature is shifted from the core center

  11. Measurement and CFD calculation of spacer loss coefficient for a tight-lattice fuel bundle

    International Nuclear Information System (INIS)

    In, Wang Kee; Shin, Chang Hwan; Kwack, Young Kyun; Lee, Chi Young

    2015-01-01

    Highlights: • Experiment and CFD analysis evaluated the pressure drop in a spacer grid. • The measurement and CFD errors for the spacer loss coefficient were estimated. • The spacer loss coefficient for the dual-cooled annular fuel bundle was determined. • The CFD prediction agrees with the measured spacer loss coefficient within 8%. - Abstract: An experiment and computational fluid dynamics (CFD) analysis were performed to evaluate the pressure drop in a spacer grid for a dual-cooled annular fuel (DCAF) bundle. The DCAF bundle for the Korean optimum power reactor (OPR1000) is a 12 × 12 tight-lattice rod array with a pitch-to-diameter ratio of 1.08 owing to a larger outer diameter of the annular fuel rod. An experiment was conducted to measure the pressure drop in spacer grid for the DCAF bundle. The test bundle is a full-size 12 × 12 rod bundle with 11 spacer grid. The test condition covers a Reynolds number range of 2 × 10 4 –2 × 10 5 by changing the temperature and flow rate of water. A CFD analysis was also performed to predict the pressure drop through a spacer grid using the full-size and partial bundle models. The pressure drop and loss coefficient of a spacer grid were predicted and compared with the experimental results. The CFD predictions of spacer pressure drop and loss coefficient agree with the measured values within 8%. The spacer loss coefficient for the DCAF bundle is estimated to be approximately 1.50 at a nominal operating condition of OPR1000, i.e., Re = 4 × 10 5

  12. In-pool damaged fuel bundle recovery

    International Nuclear Information System (INIS)

    Piascik, T.G.; Patenaude, R.S.

    1988-01-01

    While preparing to rerack the Oyster Creek Nuclear Generating Station, GPU Nuclear had need to move a damaged fuel bundle. This bundle had no upper tie plate and could not be moved in the normal manner. GPU Nuclear formed a small, dedicated project team to disassemble, package, and move this damaged bundle. The team was composed of key personnel from GPU Nuclear Fuels Projects, OCNGS Operations and Proto-Power/Bisco, a specialty contractor who has fuel bundle reconstitution and rod consolidation experience, remote tooling, underwater video systems and experienced technicians. Proven tooling, clear procedures and a simple approach were important, but the key element was the spirit of teamwork and leadership exhibited by the people involved. In spite of several emergent problems which a task of this nature presents, this small, close knit utility/vendor team completed the work on schedule and within the exposure and cost budgets

  13. Verification of heat flux and temperature calculation on the control rod outer surface

    Science.gov (United States)

    Taler, Jan; Cebula, Artur

    2011-12-01

    The paper presents heat transfer calculation results concerning a control rod of Forsmark Nuclear Power Plant (NPP). The part of the control rod, which is the object of interest, is surrounded by a mixing region of hot and cold flows and, as a consequence, is subjected to thermal fluctuations. The paper describes a numerical test which validates the method based on the solution of the inverse heat conduction problem (IHCP). The comparison of the results achieved by two methods, computational fluid dynamics (CFD) simulations and IHCP, including a description of the IHCP method used in the calculation process, shows a very good agreement between the methods.

  14. Convective heat transfer in foams under laminar flow in pipes and tube bundles.

    Science.gov (United States)

    Attia, Joseph A; McKinley, Ian M; Moreno-Magana, David; Pilon, Laurent

    2012-12-01

    The present study reports experimental data and scaling analysis for forced convection of foams and microfoams in laminar flow in circular and rectangular tubes as well as in tube bundles. Foams and microfoams are pseudoplastic (shear thinning) two-phase fluids consisting of tightly packed bubbles with diameters ranging from tens of microns to a few millimeters. They have found applications in separation processes, soil remediation, oil recovery, water treatment, food processes, as well as in fire fighting and in heat exchangers. First, aqueous solutions of surfactant Tween 20 with different concentrations were used to generate microfoams with various porosity, bubble size distribution, and rheological behavior. These different microfoams were flowed in uniformly heated circular tubes of different diameter instrumented with thermocouples. A wide range of heat fluxes and flow rates were explored. Experimental data were compared with analytical and semi-empirical expressions derived and validated for single-phase power-law fluids. These correlations were extended to two-phase foams by defining the Reynolds number based on the effective viscosity and density of microfoams. However, the local Nusselt and Prandtl numbers were defined based on the specific heat and thermal conductivity of water. Indeed, the heated wall was continuously in contact with a film of water controlling convective heat transfer to the microfoams. Overall, good agreement between experimental results and model predictions was obtained for all experimental conditions considered. Finally, the same approach was shown to be also valid for experimental data reported in the literature for laminar forced convection of microfoams in rectangular minichannels and of macrofoams across aligned and staggered tube bundles with constant wall heat flux.

  15. Study on the effect of the CANFLEX-NU fuel element bowing on the critical heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Ho Chun; Cho, Moon Sung; Jeon, Ji Su

    2001-01-01

    The effect of the CANFLEX-NU fuel element bowing on the critical heat flux is reviewed and analyzed, which is requested by KINS as the Government design licensing condition for the use of the fuel bundles in CANDU power reactors. The effect of the gap between two adjacent fuel elements on the critical heat flux and onset-of-dryout power is studied. The reduction of the width of a single inter-rod gap from its nominal size to the minimum manufacture allowance of 1 mm has a negligible effects on the thermal-hydraulic performance of the bundle for the given set of boundary conditions applied to the CANFLEX-43 element bundle in an uncrept channel. As expected, the in-reactor irradiation test results show that there are no evidence of the element bow problems on the bundle performance.

  16. CFD analysis of pressure drop across grid spacers in rod bundles compared to correlations and heavy liquid metal experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Batta, A., E-mail: batta@kit.edu; Class, A.G., E-mail: class@kit.edu

    2017-02-15

    Early studies of the flow in rod bundles with spacer grids suggest that the pressure drop can be decomposed in contributions due to flow area variations by spacer grids and frictional losses along the rods. For these shape and frictional losses simple correlations based on theoretical and experimental data have been proposed. In the OECD benchmark study LACANES it was observed that correlations could well describe the flow behavior of the heavy liquid metal loop including a rod bundle with the exception of the core region, where different experts chose different pressure-loss correlations for the losses due to spacer grids. Here, RANS–CFD simulations provided very good data compared to the experimental data. It was observed that the most commonly applied Rehme correlation underestimated the shape losses. The available correlations relate the pressure drop across a grid spacer to the relative plugging of the spacer i.e. solidity e{sub max}. More sophisticated correlations distinct between spacer grids with round or sharp leading edge shape. The purpose of this study is to (i) show that CFD is suitable to predict pressure drop across spacer grids and (ii) to access the generality of pressure drop correlations. By verification and validation of CFD results against experimental data obtained in KALLA we show (i). The generality (ii) is challenged by considering three cases which yield identical pressure drop in the correlations. First we test the effect of surface roughness, a parameter not present in the correlations. Here we compare a simulation assuming a typical surface roughness representing the experimental situation to a perfectly smooth spacer surface. Second we reverse the flow direction for the spacer grid employed in the experiments which is asymmetric. The flow direction reversal is chosen for convenience, since an asymmetric spacer grid with given blockage ratio, may result in different flow situations depending on flow direction. Obviously blockage

  17. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical 3-Rod and 7-Rod Clusters

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M; Hernborg, G; Flinta, J E

    1964-08-15

    The present report deals with measurements of burnout conditions for flow of boiling water in vertical 3-rod and 7-rod clusters. Data were obtained,in respect of heating the rods only, as well as for simultaneous uniform and non-uniform heating of the rods and the shroud. Totally, 520 runs were performed. In the case of equal heat fluxes on all surfaces of the channels, burnout always occurred on the rods, and the data were low by a factor of about 1.3 compared with round duct data. When only the rods were heated, the data showed very low burnout values in comparison with the results for total uniform heating and round ducts. This disagreement was explained by considering the climbing film flow model and the fact that only a fraction of the channel perimeter was heated. For simultaneous and non-uniform heating of the rods and the shroud it was found that the shroud could be overloaded up to 50 per cent without reducing the margin of safety in respect of burnout for the rod cluster. Finally, a correlation for predicting burnout conditions in round ducts, annuli and rod clusters has been presented. This correlation predicts the burnout heat fluxes for the present measurements and previously obtained annuli measurements within {+-} 5 per cent.

  18. Experimental studies of the effect of rod spacing on burnout in a simulated rod bundle

    International Nuclear Information System (INIS)

    Lee, D.H.; Little, R.B.

    1962-08-01

    Tests on a dumb-bell shaped flow passage simulating the gap between rods in a fuel element indicated that burnout was not significantly affected by inter-rod gap in the range 0.032'' to 0.22''. Test conditions were: 960 p.s.i.a., 2 x 10 6 1b/ft 2 hr mass velocity, and 10% mean exit quality with vertical upflow of water. (author)

  19. Investigation and application of reduced-order methods for flows study in heat exchanger tube bundles

    International Nuclear Information System (INIS)

    Pomarede, M.

    2012-01-01

    The objective of this thesis is to study the ability of model reduction for investigations of flow-induced vibrations in heat exchangers tube bundle systems.These mechanisms are a cause of major concern because heat exchangers are key elements of nuclear power plants and on-board stoke-holds.In a first part, we give a recall on heat exchangers functioning and on vibratory problems to which they are prone. Then, complete calculations leaded with the CFD numerical code Code-Saturne are carried out, first for the flow around a single circular cylinder (fixed then elastically mounted) and then for the case of a tube bundle system submitted to cross-flow. Reduced-order method POD is applied to the flow resolution with fixed structures. The obtained results show the efficiency of this technique for such configurations, using stabilization methods for the dynamical system resolution in the tube-bundle case. Multiphase-POD, which is a method enabling the adaptation of POD to fluid-structure interactions, is applied. Large displacements of a single cylinder elastically mounted under cross-flow, corresponding to the lock-in phenomenon,are well reproduced with this reduction technique. In the same way, large displacements of a confined moving tube in a bundle are shown to be faithfully reconstructed.Finally, the use of model reduction is extended to parametric studies. First, we propose to use the method which consists in projecting Navier-Stokes equations for several values of the Reynolds number on to a unique POD basis. The results obtained confirm the fact that POD predictability is limited to a range of parameter values. Then, a basis interpolation method, constructed using Grassmann manifolds and allowing the construction of a POD basis from other pre-calculated basis, is applied to basic cases. (author)

  20. Dynamic behaviour of FBR fuel pin bundles

    International Nuclear Information System (INIS)

    Martin, P.H.; Van Dorsselaere, J.P.; Ravenet, A.

    1990-01-01

    A programme of shock tests on a fast neutron reactor subassembly model (SPX1 geometry) including a complete bundle of fuel pins (dummy elements) is being carried out in the BELIER test facility at Cadarache. The purpose of these tests is: to determine the distribution of dynamic forces applied to the fuel rod clads under the impact conditions encountered in a reactor during a earthquake; to reduce as much as possible the conservatism of the methods presently used for the calculation of those forces. The test programme, now being completed, consists of the following steps: impacts on the mock-up in air with an non-compact bundle (situation of the subassembly at beginning of life (BOL) with clearances within the bundle); impacts under the same conditions but with fluid (water) in the subassembly; impacts on the mock-up in air and with a compacted bundle (simulating the conditions of an end-of-life (EOL) bundle with no clearance within the bundle). The accelerations studied in these tests cover the range encountered in design calculations for the subassembly frequencies in beam mode. (author)

  1. Thermal performance of the nuclear fuel rods submitted to angular variation of the heat exchanger coefficients

    International Nuclear Information System (INIS)

    Carvalho, A.M.M. de.

    1984-01-01

    Generally, LMFBR fuel rods consist of fuel pellets encapsulated in cladding tubes. These tubes are wrapped by a helical wire, working as a spacer. Distortions in the rod temperature distribution and in the external heat flux can be generated by angular variations in the local heat transfer coefficients due to the wire, by excentricity between pellet and clad or by ovalization of the cladding tube. Also, the temperature distributions can be affected by fuel densification, reestructuring and swelling. The present work consists of the development of a computer code in order to analyse the fuel rod performance as function of geometrical and operational effects, in steady state regime. (Author) [pt

  2. Distribution of two-phase flow thermal and hydraulic parameters over the cross-section of channels with a rod bundle

    International Nuclear Information System (INIS)

    Mironov, Yu.V.; Shpanskij, S.V.

    1975-01-01

    The paper describes PUCHOK-2, a program for thermohydraulic calculation of a channel with a bundle of smooth fuel elements. The pro.gram takes into consideration the non-uniformity of flow parameter distributions over the channel cross-section. The channel cross-section was divided into elementary cells, within which changes in flow parameters (mass velocity, heat- and steam content) were disregarded. The bundle was considered to be a system of parallel interconnected channels. Accounting for equal pressure drops in all the cells, the above model led to a system of non-linear algebraic equations. The system of equations was solved by the method of successive approximations. Theoretical results were compared with experimental data

  3. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical 3-Rod and 7-Rod Clusters

    International Nuclear Information System (INIS)

    Becker, Kurt M.; Hernborg, G.; Flinta, J.E.

    1964-08-01

    The present report deals with measurements of burnout conditions for flow of boiling water in vertical 3-rod and 7-rod clusters. Data were obtained,in respect of heating the rods only, as well as for simultaneous uniform and non-uniform heating of the rods and the shroud. Totally, 520 runs were performed. In the case of equal heat fluxes on all surfaces of the channels, burnout always occurred on the rods, and the data were low by a factor of about 1.3 compared with round duct data. When only the rods were heated, the data showed very low burnout values in comparison with the results for total uniform heating and round ducts. This disagreement was explained by considering the climbing film flow model and the fact that only a fraction of the channel perimeter was heated. For simultaneous and non-uniform heating of the rods and the shroud it was found that the shroud could be overloaded up to 50 per cent without reducing the margin of safety in respect of burnout for the rod cluster. Finally, a correlation for predicting burnout conditions in round ducts, annuli and rod clusters has been presented. This correlation predicts the burnout heat fluxes for the present measurements and previously obtained annuli measurements within ± 5 per cent

  4. Gelation And Mechanical Response of Patchy Rods

    Science.gov (United States)

    Kazem, Navid; Majidi, Carmel; Maloney, Craig

    We perform Brownian Dynamics simulations to study the gelation of suspensions of attractive, rod-like particles. We show that details of the particle-particle interactions can dramatically affect the dynamics of gelation and the structure and mechanics of the networks that form. If the attraction between the rods is perfectly smooth along their length, they will collapse into compact bundles. If the attraction is sufficiently corrugated or patchy, over time, a rigid space spanning network forms. We study the structure and mechanical properties of the networks that form as a function of the fraction of the surface that is allowed to bind. Surprisingly, the structural and mechanical properties are non-monotonic in the surface coverage. At low coverage, there are not a sufficient number of cross-linking sites to form networks. At high coverage, rods bundle and form disconnected clusters. At intermediate coverage, robust networks form. The elastic modulus and yield stress are both non-monotonic in the surface coverage. The stiffest and strongest networks show an essentially homogeneous deformation under strain with rods re-orienting along the extensional axis. Weaker, clumpy networks at high surface coverage exhibit relatively little re-orienting with strong non-affine deformation. These results suggest design strategies for tailoring surface interactions between rods to yield rigid networks with optimal properties. National Science Foundation and the Air Force Office of Scientific Research.

  5. Considerations about the utilization of electrically heated rods used for simulation of nuclear fuel pins

    International Nuclear Information System (INIS)

    Lima, R. de C.F. de; Carajilescov, P.

    1987-01-01

    The dinamic behavior of electrically heated rods used for simulation of nuclear fuel pins in nuclear power transients, is analysed by the application of the lumped parameter and the finite difference methods. Deviations of the rods surface conditions, for extreme accidental transient conditions are presented and discussed. (author) [pt

  6. Analysis of heat transfer from fuel rods with externally attached thermocouples

    International Nuclear Information System (INIS)

    Gill, C.R.; Coddington, P.

    1988-05-01

    This paper describes the development of 2 and 3 dimensional finite element heat conduction models to simulate the behaviour of the external thermocouples attached to the LOFT fuel rods during the blowdown phase of a large break loss-of-coolant accident. To establish the model and determine the thermal coupling between the thermocouple and the fuel rod extensive use was made of two series of experiments performed at INEL in the LOFT Test Support Facility (LTSF). These experiments were high pressure reflood experiments with fluid conditions 'typical' of those seen during the bottom-up flow period of the LOFT experiments. (author)

  7. Pressure drop redistribution experimental analysis in axial flow along the bundles

    International Nuclear Information System (INIS)

    Bastos Franco, C. de; Carajilescov, P.

    1992-01-01

    Fuel elements of PWR type nuclear reactors are composed of rod bundles, arranged in square arrays, held by grid type spacers. The coolant flows axially along the bundle. Although such elements are laterally open, pressure drop experiments are performed in closed type test sections, originating the appearance of subchannels of different geometries. Utilizing a test section of two bundles of 4 x 4 pins and performing experiments with and without separation between the bundles, the flow redistribution factors, the friction, and the grid drag coefficients were determined for the interior subchannels. 03 refs, 06 figs, 02 tabs. (B.C.A.)

  8. Predictions of Critical Heat Flux Using the ASSERT-PV Subchannel Code for a CANFLEX Variant Bundle

    International Nuclear Information System (INIS)

    Onder, Ebru Nihan; Leung, Laurence; Kim, Hung; Rao, Yanfei

    2009-01-01

    The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced CANDU 1 reactor fuel bundle. Based primarily on the CANFLEX 2 fuel bundle, several geometry changes (such as element sizes and pitchcircle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures

  9. Analyzing the BWR rod drop accident in high-burnup cores

    International Nuclear Information System (INIS)

    Diamond, D.J.; Neymotin, L.; Kohut, P.

    1995-01-01

    This study was undertaken for the US Nuclear Regulatory Commission to determine the fuel enthalpy during a rod drop accident (RDA) for cores with high burnup fuel. The calculations were done with the RAMONA-4B code which models the core with 3-dimensional neutron kinetics and multiple parallel coolant channels. The calculations were done with a model for a BWR/4 with fuel bundles having burnups up to 30 GWd/t and also with a model with bundle burnups to 60 GWd/t. This paper also discusses potential sources of uncertainty in calculations with high burnup fuel. One source is the ''rim'' effect which is the extra large peaking of the power distribution at the surface of the pellet. This increases the uncertainty in reactor physics and heat conduction models that assume that the energy deposition has a less peaked spatial distribution. Two other sources of uncertainty are the result of the delayed neutron fraction decreasing with burnup and the positive moderator temperature feedback increasing with burnup. Since these effects tend to increase the severity of the event, an RDA calculation for high burnup fuel will underpredict the fuel enthalpy if the effects are not properly taken into account. Other sources of uncertainty that are important come from the initial conditions chosen for the RDA. This includes the initial control rod pattern as well as the initial thermal-hydraulic conditions

  10. Post test investigation of the bundle test ESBU-1

    International Nuclear Information System (INIS)

    Hagen, S.; Kapulla, H.; Malauschek, H.; Wallenfels, K.P.; Buescher, B.J.

    1986-08-01

    This KfK report describes the post test investigation of bundle experiment ESBU-1. ESBU-1 was the first of two bundle tests on the temperature escalation of Zircaloy clad fuel rods. The investigation of the temperature escalation is part of the program of out-of-pile experiments performed within the frame work of the PNS - Severe Fuel Damage program. The bundle was composed of a 3x3 fuel rod array of our fuel rod simulators (control tungsten heater, UO 2 -ring pellet and Zircaloy cladding). The length was 0.4 meter. After the test the bundle was embedded in epoxy and cut by a diamant saw. The cross sections are investigated by metallographic, SEM and EMP examinations. The results of these examinations are in good agreement with the seperate effects tests investigation of the PNS SFD-Program and inpile experiments of the Power Burst Facility. The investigations show that liquid Zircaloy dissolves UO 2 by taking away the oxygen from the oxide. Depending on the overall oxygen content the (U,Zr,O)-melt forms at refreezing a) three phases (low oxygen content): metallic α-Zry(U), a uranium-rich metallic (U,Zr)alloy, and a (U,Zr)O 2 mixed oxide, or b) two phases (high oxygen content): α-Zr(O) and the (U,Zr)O 2 mixed oxide. c) In melt regions where the local oxidation was very severe, such as in steam contact, only the (U,Zr)O 2 mixed oxide is formed already at test temperature. Also ZrO 2 formed during the initial time of the test is dissolved by the melt. (orig.) [de

  11. Numerical prediction on turbulent heat transfer of a spacer ribbed fuel rod for high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Takase, Kazuyuki

    1994-11-01

    The turbulent heat transfer of a fuel rod with three-dimensional trapezoidal spacer ribs for high temperature gas-cooled reactors was analyzed numerically using the k-ε turbulence model, and investigated experimentally using a simulated fuel rod under the helium gas condition of a maximum outlet temperature of 1000degC and pressure of 4MPa. From the experimental results, it found that the turbulent heat transfer coefficients of the fuel rod were 18 to 80% higher than those of a concentric smooth annulus at a region of Reynolds number exceeding 2000. On the other hand, the predicted average Nusselt number of the fuel rod agreed well with the heat transfer correlation obtained from the experimental data within a relative error of 10% with Reynolds number of more than 5000. It was verified that the numerical analysis results had sufficient accuracy. Furthermore, the numerical prediction could clarify quantitatively the effects of the heat transfer augmentation by the spacer rib and the axial velocity increase due to a reduction in the annular channel cross-section. (author)

  12. Numerical analysis on the condensation heat transfer and pressure drop characteristics of the horizontal tubes of modular shell and tube-bundle heat exchanger

    International Nuclear Information System (INIS)

    Ko, Seung Hwan; Park, Hyung Gyu; Kim, Charn Jung; Park, Byung Kyu

    2001-01-01

    A numerical analysis of the heat and mass transfer and pressure drop characteristics in modular shell and tube bundle heat exchanger was carried out. Finite concept method based on FVM and κ-ε turbulent model were used for this analysis. Condensation heat transfer enhanced total heat transfer rate 4∼8% higher than that of dry heat exchanger. With increasing humid air inlet velocity, temperature and relative humidity, and with decreasing heat exchanger aspect ratio and cooling water velocity, total heat and mass transfer rate could be increased. Cooling water inlet velocity had little effect on total heat transfer

  13. Core conversion from rod to plate type fuel elements in research reactors

    International Nuclear Information System (INIS)

    Khattab, M.S.; Mina, A.R.

    1997-01-01

    Core thermalhydraulic analysis have been performed for rod and plate types fuel elements without altering the core bundles square grid spacer (68 mm, side) and coolant mass flow rate. The U O 2 -Mg, 10% enrichment rod type fuel elements are replaced by the MTR plate type, U-Al alloy of 20% enrichment. Coolant mass flux increased from 2000 kg/m 2 S to 5000 kg/m 2 S. Reactor power could be upgraded from 2 to 10 MW without significantly altering the steady state, thermal-hydraulic safety margins. Fuel, clad and coolant transient temperatures are determined inside the core hot channel during flow coast down using paret code. Residual heat removal system of 20% coolant capacity is necessary for upgrading reactor power to encounter the case of pumps off at 10 MW nominal operation. 6 figs., 2 tabs

  14. RELAP5 analysis of reflux condensation behavior in heat transfer tube bundle of a steam generator

    International Nuclear Information System (INIS)

    Minami, Noritoshi; Chikusa, Toshiaki; Nagae, Takashi; Murase, Michio

    2007-01-01

    In case of loss of the residual heat removal system and other alternative cooling methods under mid-loop operation during shutdown of the pressurized water reactor plant, reflux condensation in the steam generator (SG) may be an effective heat removal mechanism. In reflux condensation experiments 7.2c with injection of nitrogen gas using the BETHSY facility in France, which is a scale model of a pressurized water reactor plant, 34 heat transfer tubes were divided into two kinds of flow patterns, which were steam forward flow and nitrogen reverse flow. In this study, we simulated the BETHSY experiments using the transient analysis code RELAP5. Modifying calculation equations for interfacial friction force and wall friction force between the inlet plenum and heat transfer tubes, nitrogen reverse flow was successfully simulated. In calculations with alteration of the flow area ratio to two flow channels for the heat transfer tube bundle, the number of active tubes with the maximum nitrogen recirculation flow rate agreed rather well with the observed number of active tubes. In calculations with three flow channels for the heat transfer tube bundle, the average number of active tubes in several calculations with different flow area ratios of the three flow channels predicted the number of active tubes well. (author)

  15. BRIGITTE, Dose Rate and Heat Source and Energy Flux for Self-Absorbing Rods

    International Nuclear Information System (INIS)

    Jegu, M.; Clement, M.

    1978-01-01

    1 - Nature of physical problem solved: Calculation of dose rate, heat sources or energy flux. The sources are self-absorbing radioactive rods. The shielding consists of blocks of which the cross section can be defined. 2 - Method of solution: Exponential attenuation and build-up factor between source points and detector points. Source integration with error estimate. Automatic or controlled build-up with monitor print-out. 3 - Restrictions on the complexity of the problem: Number of energy points, regions, detector points, abscissa points of the rod, vertical position of the rod, are all limited to ten. The maximum total number of vertical steps is 124

  16. QUENCH-LOCA program at KIT and results of the QUENCH-L0 bundle test

    International Nuclear Information System (INIS)

    Stuckert, J.; Grosse, M.; Roessger, C.; Steinbrueck, M.; Walter, M.

    2012-01-01

    of Technology (KIT) in the QUENCH facility. Compared to single-rod experiments, bundle tests have the advantage of studying the mutual interference of rod ballooning among fuel rod simulators as well as the local coolant channel blockages in a more realistic arrangement. The program was started in 2010 with the QUENCH-L0 scoping test used 21 heated rods with as-received Zry-4 claddings. Then advanced claddings (Duplex D4, M5 registered , ZIRLO trademark) will be used. (orig.)

  17. QUENCH-LOCA program at KIT and results of the QUENCH-L0 bundle test

    Energy Technology Data Exchange (ETDEWEB)

    Stuckert, J.; Grosse, M.; Roessger, C.; Steinbrueck, M.; Walter, M. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

    2012-11-01

    Karlsruhe Institute of Technology (KIT) in the QUENCH facility. Compared to single-rod experiments, bundle tests have the advantage of studying the mutual interference of rod ballooning among fuel rod simulators as well as the local coolant channel blockages in a more realistic arrangement. The program was started in 2010 with the QUENCH-L0 scoping test used 21 heated rods with as-received Zry-4 claddings. Then advanced claddings (Duplex D4, M5 {sup registered}, ZIRLO trademark) will be used. (orig.)

  18. Determination of transient temperature and heat flux on the surface of a reactor control rod based on temperature measurements at the interior points

    International Nuclear Information System (INIS)

    Cebula, Artur; Taler, Jan

    2014-01-01

    The paper presents heat transfer calculation results concerning a control rod of nuclear power plant. Apart from numerical calculation results, experimental heat transfer measurements of the control rod model are also presented. The control rod that is the object of interest is surrounded by a mixing region of hot and cold streams and, as a consequence, is subjected to thermal fluctuations. The paper describes a method based on the solution of the inverse heat conduction problem (IHCP) for determining heat flux on the outer surface of the rod. Numerical tests were conducted to validate the method by comparison of the results with the time changes of surface temperature and heat flux which were obtained from the computational fluid dynamics (CFD) simulation of the mixing process. A measuring instrument was designed to measure the heat flux at the outer surface of the control rod model. In addition, the principle of operation and construction of heat flux meter is presented in detail. -- Highlights: • Temperature and heat flux estimation during cooling of control rod are presented. • The inverse technique is based on the space marching method. • The instrument for surface heat flux measurement was manufactured and tested. • CFD simulations were used to validate the developed inverse technique. • Actual data were used to demonstrate practical applicability of the method

  19. Fuel Rod Vibration Measurement Method using a Flap and its Verification

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Joo Young; Park, Nam Gyu; Suh, Jung Min; Jeon, Kyeong Lak [KEPCO NF Co., Daejeon (Korea, Republic of)

    2011-10-15

    Flow-induced vibration is a critical factor for the mechanical integrity of a fuel rod. This vibration can cause leaked fuel through the mechanism, such as grid to rod fretting. To minimize the failures caused by flow-induced vibration, a robust design is needed which takes into account vibrational characteristics. That is, the spacer grid design should be developed to avoid any excessive vibration. On the one hand, if fuel rod vibration can be measured, an estimation of the excitation forces, which are a critical cause of rod failure, should be possible. Therefore, by applying an external force, flow-induced vibration can be roughly estimated when the fuel rod vibration model is used. KEPCO Nuclear Fuel developed the test loop to research flow-induced vibration as shown in Fig.1. The investigation flow-induced vibration (INFINIT) - the test facility - can measure the grid strap vibration and pressure drop of a 5x5 small scale fuel bundle. Basically, using a Laser Doppler Vibrometer (LDV), the vibration of a structure immersed in high speed fluid can be measured. Grid strap vibration is easily measured using an LDV. However, it is quite difficult to measure fuel rod vibration because of the round surface shape of the rods. In addition, measuring current method using the LDV, it was only possible to directly measure fuel rod vibration at the first row of the bundle as the rods behind the first row are obscured. To solve this problem, a thin flap, as shown in Fig. 2(a) can be used as a reflecting target, gaining access to rods within the bundle. The flap is attached to the fuel rod, as in Fig. 2(b). As a result, most of the inner rod vibration can be measured. Before using a flap to measure fuel rod vibration, a verification process was needed to show whether the LDV signal from the flap vibration provided equivalent and reliable signals. Therefore, impact testing was carried out on the fuel rod using a flap. The LDV signals were then compared with accelerometer

  20. Pressure drop variation as a function of axial and radial power distribution in CANDU fuel channel with standard and CANFLEX 43 bundles

    International Nuclear Information System (INIS)

    Catana, Alexandru; Department of Energy Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

    2007-01-01

    CANDU 600 nuclear reactors are usually fuelled with STANDARD (STD), 37 rods fuel bundles. Natural uranium (NU) dioxide (UO 2 ), is used as fuel composition. A new fuel bundle geometry called CANFLEX (CFX) with 43 rods is proposed and some new fuel composition are considered. Flexibility is the key word for the attempt to use some different fuel geometries and compositions for CANDU 600 nuclear reactors as well as for innovative ACR-700/1000 nuclear reactors. The fuel bundle considered in this paper is CFX-RU-0.90 that encodes the CANFLEX geometry, recycled dioxide uranium (RU) with 0.90% enrichment. The goal of this proposal is ambitious: a higher average discharge burn-up up to 14000 MWd/tU and, for the same amount of generated electric power, reduction in nuclear fuel fabrication, reduction of spent nuclear fuel radioactive waste and reduction of refueling operational work by using fewer bundles. An improved sub-channel approach for thermal-hydraulic analysis is used in this paper to compute some flow parameters, mainly the pressure drop along the CANDU 600 fuel channel when STD or CFX-RU-0.90 fuel bundles. Also an intermediate CFX-NU fuel bundle are used, for gradual comparison. For CFX-RU- 0.90 four fuel bundle shift refueling scheme is used instead of eight, that will determine different axial power distributions. At the same time radial power distribution is affected by the geometry and by the fuel composition of fuel bundle type used. Some other thermal-hydraulic flow parameters will be influenced, too. One of the most important parameter is pressure drop (PD) along the fuel channel because of its importance in drag force evaluation. We start with an axial power distribution, which is characteristic for a refueling scheme of eight or four fuel bundles on a shift. Comparative results are presented between STD37, CFX-NU CFX-RU-0.90 fuel bundles in a CANDU nuclear reactor operating conditions. Neutron flux distribution analysis shows that four bundle shift

  1. Thermohydraulics in rod bundles and critical heat flux in transient conditions in a tube

    International Nuclear Information System (INIS)

    Courtaud, M.; Roumy, R.

    1975-01-01

    After the determination of the scaling factor of Stevens's similitude for the pressure range of pressurized water vectors by comparison of critical heat flux data in from and in water, some examples of studies performed with freon are shown. The efficiency of the mixing vanes of spacer grids has been determined on the mixing phenomenon in single phase on critical heat flux. A calculation performed with the code FLICA using subchannel analysis on freon data transposed in water is in good agreement with the experiment. The influence of the number of spacer grids has been also shown. Critical heat fluxes have been determined in water at 140 bar in steady state and transient conditions on two tubular test sections. During the transient tests the flow rate was reduced by half in 0.5 seconds and the reincreased heat flux and inlet temperature remaining constant. These tests have shown the validity of the method which consists in using a critical heat flux correlation determined in steady state conditions applied with local transient conditions of enthalpy and mass velocity computed with the FLICA code [fr

  2. Drive-in device for long thin rods into narrow cavitations, especially for control-shutdown rods e.g. of nuclear reactors

    International Nuclear Information System (INIS)

    Flessner, H.; Paeserack, U.

    1974-01-01

    The auxiliary device serves as holder for long and thin rods, e.g. control rods, transported hanging in bundles, when these are lowered into narrow cavities. It is constructed as a rod grab vertically movable at the end of a guide tube. A comb-shaped trap in connection with a guide rod serves for lateral support of the lower ends of the rods hanging on the grab. This guide rod can be moved in vertical direction by means of two pairs of convex rollers resting on the inner guide tube. In addition, the guide rod has a prolongation carrying a traverse by means of an abutment on the lower end. With these auxiliaries amongst others, the trap can be brought into a horizontal position by turning around an axis with the control rods meshing with the teeth of the trap while the parallelism of the rods is kept up during transport. (DG) [de

  3. THEBES: a thermal hydraulic code for the calculation of transient two phase flow in bundle geometry

    International Nuclear Information System (INIS)

    Camous, F.

    1983-01-01

    The three dimensional thermal hydraulic code THEBES, capable to calculate transient boiling of sodium in rod bundles is described here. THEBES, derived from the transient single phase code SABRE-2A, was developed in CADARACHE by the SIES to analyse the SCARABEE N loss of flow experiments. This paper also presents the results of tests which were performed against various types of experiments: (1) transient boiling in a 7 pin bundle simulating a partial blockage at the bottom of a subassembly (rapid transient SCARABEE 7.2 experiment), (2) transient boiling in a 7 pin bundle simulating a coolant coast down (slow transient SCARABEE 7.3 experiment), (3) steady local and generalised boiling in a 19 pin bundle (GR 19 I experiment), (4) transient boiling in a 19 pin bundle simulating a coolant coast down (GR 19 I experiment), (5) steady local boiling in a 37 pin bundle with internal blockage (MOL 7C experiment). Excellent agreement was found between calculated and experimental results for these different situations. Our conclusion is that THEBES is able to calculate transient boiling of sodium in rod bundles in a quite satisfying way

  4. Research of heat transfer of staggered horizontal bundles of finned tubes at free air convection

    Science.gov (United States)

    Novozhilova, A. V.; Maryna, Z. G.; Samorodov, A. V.; Lvov, E. A.

    2017-11-01

    The study of free-convective processes is important because of the cooling problem in many machines and systems, where other ways of cooling are impossible or impractical. Natural convective processes are common in the steam turbine air condensers of electric power plants located within the city limits, in dry cooling towers of circulating water systems, in condensers cooled by air and water, in radiators cooling oil of power electric transformers, in emergency cooling systems of nuclear reactors, in solar power, as well as in air-cooling of power semiconductor energy converters. All this makes actual the synthesis of the results of theoretical and experimental research of free convection for heat exchangers with finned tube bundles. The results of the study of free-convection heat transfer for two-, three- and four-row staggered horizontal bundles of industrial bimetallic finned tubes with finning factor of 16.8 and equilateral tubes arrangement are presented. Cross and diagonal steps in the bundles are the same: 58; 61; 64; 70; 76; 86; 100 mm, which corresponds to the relative steps: 1.042; 1.096; 1.152; 1.258; 1.366; 1.545; 1.797. These steps are standardized for air coolers. An equation for calculating the free-convection heat transfer, taking into account the influence of geometrical parameters in the range of Rayleigh number from 30,000 to 350,000 with an average deviation of ± 4.8%, has been obtained. The relationship presented in the article allows designing a wide range of air coolers for various applications, working in the free convection modes.

  5. Assessment of boiling transition analysis code against data from NUPEC BWR full-size fine-mesh bundle tests

    International Nuclear Information System (INIS)

    Utsuno, Hideaki; Ishida, Naoyuki; Masuhara, Yasuhiro; Kasahara, Fumio

    2004-01-01

    Transient BT analysis code TCAPE based on mechanistic methods coupled with subchannel analysis has been developed for the evaluation on fuel integrity under abnormal operations in BWR. TCAPE consisted mainly of the drift-flux model, the cross-flow model, the film model and the heat transfer model. Assessment of TCAPE has been performed against data from BWR full-size fine-mesh bundle tests (BFBT), which consisted of two major parts: the void distribution measurement and the critical power measurement. Code and data comparison was made for void distributions with varying number of unheated rods in simulated actual fuel assembly. Prediction of steady-state critical power was compared with the measurement on full-scale bundle under a range of BWR operational conditions. Although the cross-sectional averaged void fraction was underestimated when it became lower, the accuracy was obtained that the averaged ratio 0.910 and its standard deviation 0.076. The prediction of steady-state critical power agreed well with the data in the range of BWR operations, where the prediction accuracy was obtained that the averaged ratio 0.997 and its standard deviation 0.043. These results demonstrated that TCAPE is well capable to predict two-phase flow distribution and liquid film dryout phenomena occurring in BWR rod bundles. Part of NUPEC BFBT database will be made available for an international benchmark exercise. The code assessment shall be continued against the OECD/NRC benchmark based on BFBT database. (author)

  6. Measurements of Sheath Temperature Profiles in Bruce LVRF Bundles Under Post-Dryout Heat Transfer Conditions in Freon

    International Nuclear Information System (INIS)

    Guo, Y.; Bullock, D.E.; Pioro, I.L.; Martin, J.

    2006-01-01

    An experimental program has been completed to study the behaviour of sheath wall temperatures in the Bruce Power Station Low Void Reactivity Fuel (shortened hereafter to Bruce LVRF) bundles under post-dryout (PDO) heat-transfer conditions. The experiment was conducted with an electrically heated simulator of a string of nine Bruce LVRF bundles, installed in the MR-3 Freon heat transfer loop at the Chalk River Laboratories (CRL), Atomic Energy of Canada Limited (AECL). The loop used Freon R-134a as a coolant to simulate typical flow conditions in CANDU R nuclear power stations. The simulator had an axially uniform heat flux profile. Two radial heat flux profiles were tested: a fresh Bruce LVRF profile and a fresh natural uranium (NU) profile. For a given set of flow conditions, the channel power was set above the critical power to achieve dryout, while heater-element wall temperatures were recorded at various overpower levels using sliding thermocouples. The maximum experimental overpower achieved was 64%. For the conditions tested, the results showed that initial dryout occurred at an inner-ring element at low flows and an outer-ring element facing internal subchannels at high flows. Dry-patches (regions of dryout) spread with increasing channel power; maximum wall temperatures were observed at the downstream end of the simulator, and immediately upstream of the mid-bundle spacer plane. In general, maximum wall temperatures were observed at the outer-ring elements facing the internal subchannels. The maximum water-equivalent temperature obtained in the test, at an overpower level of 64%, was significantly below the acceptable maximum temperature, indicating that the integrity of the Bruce LVRF will be maintained at PDO conditions. Therefore, the Bruce LVRF exhibits good PDO heat transfer performance. (authors)

  7. The design of the DUPIC spent fuel bundle counter

    International Nuclear Information System (INIS)

    Menlove, H.O.; Rinard, P.M.; Kroncke, K.E.; Lee, Y.G.

    1997-05-01

    A neutron coincidence detector had been designed to measure the amount of curium in the fuel bundles and associated process samples used in the direct use of plutonium in Canadian deuterium-uranium (CANDU) fuel cycle. All of the sample categories are highly radioactive from the fission products contained in the pressurized water reactor (PWR) spent fuel feed stock. Substantial shielding is required to protect the He-3 detectors from the intense gamma rays. The Monte Carlo neutron and photon calculational code has been used to design the counter with a uniform response profile along the length of the CANDU-type fuel bundle. Other samples, including cut PWR rods, process powder, waste, and finished rods, can be measured in the system. This report describes the performance characteristics of the counter and support electronics. 3 refs., 23 figs., 6 tabs

  8. Polyelectrolyte Bundles: Finite size at thermodynamic equilibrium?

    Science.gov (United States)

    Sayar, Mehmet

    2005-03-01

    Experimental observation of finite size aggregates formed by polyelectrolytes such as DNA and F-actin, as well as synthetic polymers like poly(p-phenylene), has created a lot of attention in recent years. Here, bundle formation in rigid rod-like polyelectrolytes is studied via computer simulations. For the case of hydrophobically modified polyelectrolytes finite size bundles are observed even in the presence of only monovalent counterions. Furthermore, in the absence of a hydrophobic backbone, we have also observed formation of finite size aggregates via multivalent counterion condensation. The size distribution of such aggregates and the stability is analyzed in this study.

  9. CFD analysis of liquid metal cooled rod assembly

    Energy Technology Data Exchange (ETDEWEB)

    Son, H.M.; Suh, K.Y. [Seoul National Univ. (Korea, Republic of)

    2007-07-01

    The model subassembly of the BREST-type reactor core is a pin bundle of square arrangement. In this bundle there are two zones which differ with respect to pin diameters and level of heat production. The model pin bundle contains one spacer grid which is located near the midplane of the rod bundle geometry. The coolant consists of a eutectic alloy of 22% sodium (Na) plus 78% potassium (K). Experiments were performed in order to observe the thermal hydraulic behavior of the liquid metal coolant in the BREST core simulator. Results were obtained for the coolant exit temperatures, central measuring pin simulator external surface temperatures, and coolant velocities at the perimeter of the measuring pin simulator. A computational fluid dynamics (CFD) code is used to simulate the liquid metal flows in subchannels. Semi-fine mesh structures were used to model the flow with reasonable accuracy and speed once rigorous node resolution dependency had been tested. A subchannel analysis code was used to investigate the flows as well. Since the subchannel analysis code is based on a lumped parameter model, it only calculates the subchannel averaged velocity values. The CFD code results were averaged on the subchannel basis to be comparable with the results from the subchannel code. The mixing vane is not considered for the time being so as to simplify the problem and to reduce the computational cost. The two codes showed similar results. The difference between the experimental and computational results is considered to mainly originate from the existence of the mixing vane. (authors)

  10. CFD analysis of liquid metal cooled rod assembly

    International Nuclear Information System (INIS)

    Son, H.M.; Suh, K.Y.

    2007-01-01

    The model subassembly of the BREST-type reactor core is a pin bundle of square arrangement. In this bundle there are two zones which differ with respect to pin diameters and level of heat production. The model pin bundle contains one spacer grid which is located near the midplane of the rod bundle geometry. The coolant consists of a eutectic alloy of 22% sodium (Na) plus 78% potassium (K). Experiments were performed in order to observe the thermal hydraulic behavior of the liquid metal coolant in the BREST core simulator. Results were obtained for the coolant exit temperatures, central measuring pin simulator external surface temperatures, and coolant velocities at the perimeter of the measuring pin simulator. A computational fluid dynamics (CFD) code is used to simulate the liquid metal flows in subchannels. Semi-fine mesh structures were used to model the flow with reasonable accuracy and speed once rigorous node resolution dependency had been tested. A subchannel analysis code was used to investigate the flows as well. Since the subchannel analysis code is based on a lumped parameter model, it only calculates the subchannel averaged velocity values. The CFD code results were averaged on the subchannel basis to be comparable with the results from the subchannel code. The mixing vane is not considered for the time being so as to simplify the problem and to reduce the computational cost. The two codes showed similar results. The difference between the experimental and computational results is considered to mainly originate from the existence of the mixing vane. (authors)

  11. The effects of off-center pellets on the temperature distribution and the heat flux distribution of fuel rods in nuclear reactors

    International Nuclear Information System (INIS)

    Peng Muzhang; Xing Jianhua

    1986-01-01

    This paper analyzes the effects of off-center pellets on the steady state temperature distribution and heat flux distribution of fuel rods in the nuclear reactors, and derives the dimensionless temperature distribution relationships and the dimensionless heat flux distribution relationship from the fuel rods with off-center pellets. The calculated results show that the effects of off-center will result in not only deviations of the highest temperature placement in the fuel pellets, but also the circumferentially nonuniform distributions of the temperatures and heat fluxes of the fuel rod surfaces

  12. Heat Removal Performance of Hybrid Control Rod for Passive In-Core Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-10-15

    The two-phase closed heat transfer device can be divided by thermosyphon heat pipe and capillary wicked heat pipe which uses gravitational force or capillary pumping pressure as a driving force of the convection of working fluid. If there is a temperature difference between reactor core and ultimate heat sink, the decay heat removal and reactor shutdown is possible at any accident conditions without external power sources. To apply the hybrid control rod to the commercial nuclear power plants, its modelling about various parameters is the most important work. Also, its unique geometry is coexistence of neutron absorber material and working fluid in a cladding material having annular vapor path. Although thermosyphon heat pipe (THP) or wicked heat pipe (WHP) shows high heat transfer coefficients for limited space, the maximum heat removal capacity is restricted by several phenomena due to their unique heat transfer mechanism. Validation of the existing correlations on the annular vapor path thermosyphon (ATHP) which has different wetted perimeter and heated diameter must be conducted. The effect of inner structure, and fill ratio of the working fluid on the thermal performance of heat pipe has not been investigated. As a first step of the development of hybrid heat pipe, the ATHP which contains neutron absorber in the concentric thermosyphon (CTHP) was prepared and the thermal performance of the annular thermosyphon was experimentally studied. The heat transfer characteristics and flooding limit of the annular vapor path thermosyphon was studied experimentally to model the performance of hybrid control rod. The following results were obtained: (1) The annular vapor path thermosyphon showed better evaporation heat transfer due to the enhanced convection between adiabatic and condenser section. (2) Effect of fill ratio on the heat transfer characteristics was negligible. (3) Existing correlations about flooding limit of thermosyphon could not reflect the annular vapor

  13. Impact of bundle deformation on CHF: ASSERT-PV assessment of extended burnup Bruce B bundle G85159W

    International Nuclear Information System (INIS)

    Rao, Y.F.; Manzer, A.M.

    2005-01-01

    This paper presents a subchannel thermalhydraulic analysis of the effect on critical heat flux (CHF) of bundle deformation such as element bow and diametral creep. The bundle geometry is based on the post-irradiation examination (PIE) data of a single bundle from the Bruce B Nuclear Generating Station, Bruce B bundle G85159W, which was irradiated for more than two years in the core during reactor commissioning. The subchannel code ASSERT-PV IST is used to assess changes in CHF and dryout power due to bundle deformation, compared to the reference, undeformed bundle. (author)

  14. Development of electrically heated rods with resistive element of graphite or carbon/carbon composites for simulating transients in nuclear reactors

    International Nuclear Information System (INIS)

    Polidoro, H.A.

    1987-01-01

    Thermo-hydraulic problems, in nuclear plants are normally analysed by the use of electrically heated rods. The direct or indirect heater rods are limited in their use because, for high temperatures and high heat flux, the heating element temperature approach its melting point. The use of platinum or tantalum is not economically viable. Graphite and carbon/carbon composites are alternative materials because they are good electrical conductors and have good mechanical properties at high temperatures. Graphite and carbon/carbon composites were used to make heating elements for testing by indirect heating. The swaging process used to reduce the cladding diameter prevented the fabrication of graphite heater rods. Carbon/carbon composite used to make heating elements gave good results up to a heat flux of 100 W/cm 2 . It is easy to verify that this value can be exceeded if the choice of the complementary materials for insulator and cladding improved. (author) [pt

  15. Non-Fourier Vernotte-Cattaneo numerical model for heat conduction in a BWR fuel rod

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa-Martinez, E.G.; Vazquez-Rodriguez, A.; Varela-Ham, J.R.; Espinosa-Paredes, G., E-mail: gepe@xanum.uam.mx [Universidad Autonoma Metropolitana, Area de Ingenieria en Recursos Energeticos, Iztapalapa (Mexico)

    2014-07-01

    A fuel rod mathematical model based on transient heat conduction as constitutive Non-Fourier law for Light Water Reactors (LWRs) transient analysis is presented. The structure of the fuel pellet is affected due to high temperatures and irradiation, which eventually produce fracture or cracks. In principle the fractures are saturated of gas. Then, the Fourier law of the heat conduction is not strictly applicable to describe these phenomena, where the physical properties such as thermal conductivity, heat capacity and density correspond to a heterogeneous material due to gas, and therefore the thermal diffusion process due to molecular transport in the fuel pellet is affected. From the point of view of nuclear reactor safety analysis, the heat transfer from the fuel to the coolant is crucial and superheating of the wall can cause the cladding failure. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The Non-Fourier approach presented in this work eliminates the assumption of an infinite thermal wave speed, therefore time-dependent heat sources were considered in the fuel rod heat transfer model. The numerical experiments in a BWR, show that the Non-Fourier approach is crucial in the pressurization transients such as turbine trip and reactor isolation. (author)

  16. Non-Fourier Vernotte-Cattaneo numerical model for heat conduction in a BWR fuel rod

    International Nuclear Information System (INIS)

    Espinosa-Martinez, E.G.; Vazquez-Rodriguez, A.; Varela-Ham, J.R.; Espinosa-Paredes, G.

    2014-01-01

    A fuel rod mathematical model based on transient heat conduction as constitutive Non-Fourier law for Light Water Reactors (LWRs) transient analysis is presented. The structure of the fuel pellet is affected due to high temperatures and irradiation, which eventually produce fracture or cracks. In principle the fractures are saturated of gas. Then, the Fourier law of the heat conduction is not strictly applicable to describe these phenomena, where the physical properties such as thermal conductivity, heat capacity and density correspond to a heterogeneous material due to gas, and therefore the thermal diffusion process due to molecular transport in the fuel pellet is affected. From the point of view of nuclear reactor safety analysis, the heat transfer from the fuel to the coolant is crucial and superheating of the wall can cause the cladding failure. In the classical theory of diffusion, the Fourier law of heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The Non-Fourier approach presented in this work eliminates the assumption of an infinite thermal wave speed, therefore time-dependent heat sources were considered in the fuel rod heat transfer model. The numerical experiments in a BWR, show that the Non-Fourier approach is crucial in the pressurization transients such as turbine trip and reactor isolation. (author)

  17. CFD analyses of the rod bowing effect on the subchannel outlet temperature distribution

    Energy Technology Data Exchange (ETDEWEB)

    Ekstroem, Karoliina; Toppila, Timo [Fortum Power and Heat, Fortum (Finland)

    2017-09-15

    In the Loviisa 1 and 2 nuclear power plants the subcooling margin of the hottest subchannel of the fuel assembly is monitored. The temperature of the coolant in the hottest subchannel is limited to the constant saturation temperature. Bending of the fuel rods occurs during normal operation due to the differences in the heat profiles of the rods. The coolant temperature will rise more in the subchannel with smaller flow area due to the bending and this has to be taken into account in the safety margin of subchannel enthalpy rise. Computational Fluid Dynamics (CFD) simulations are used to estimate how much the estimated maximum bow of a rod affects the temperature rise of the subchannel. The quantitative uncertainty of the predicted enthalpy rise in fuel bundle subchannel is estimated based on the uncertainty of modelling of mixing between subchannels. The measured turbulence quantities from LDA measurements of cold test assembly made in 1990s in Fortum are compared with CFD results to give uncertainty estimation for turbulence, which is further used for uncertainty estimation of mixing and simulated subchannel enthalpy rise.

  18. International Benchmark on Pressurised Water Reactor Sub-channel and Bundle Tests. Volume II: Benchmark Results of Phase I: Void Distribution

    International Nuclear Information System (INIS)

    Rubin, Adam; Avramova, Maria; Velazquez-Lozada, Alexander

    2016-03-01

    This report summarised the first phase of the Nuclear Energy Agency (NEA) and the US Nuclear Regulatory Commission Benchmark based on NUPEC PWR Sub-channel and Bundle Tests (PSBT), which was intended to provide data for the verification of void distribution models in participants' codes. This phase was composed of four exercises; Exercise 1: steady-state single sub-channel benchmark, Exercise 2: steady-state rod bundle benchmark, Exercise 3: transient rod bundle benchmark and Exercise 4: a pressure drop benchmark. The experimental data provided to the participants of this benchmark is from a series of void measurement tests using full-size mock-up tests for both Boiling Water Reactors (BWRs) and Pressurised Water Reactors (PWRs). These tests were performed from 1987 to 1995 by the Nuclear Power Engineering Corporation (NUPEC) in Japan and made available by the Japan Nuclear Energy Safety Organisation (JNES) for the purposes of this benchmark, which was organised by Pennsylvania State University. Twenty-one institutions from nine countries participated in this benchmark. Seventeen different computer codes were used in Exercises 1, 2, 3 and 4. Among the computer codes were porous media, sub-channel, systems thermal-hydraulic code and Computational Fluid Dynamics (CFD) codes. It was observed that the codes tended to overpredict the thermal equilibrium quality at lower elevations and under predict it at higher elevations. There was also a tendency to overpredict void fraction at lower elevations and underpredict it at high elevations for the bundle test cases. The overprediction of void fraction at low elevations is likely caused by the x-ray densitometer measurement method used. Under sub-cooled boiling conditions, the voids accumulate at heated surfaces (and are therefore not seen in the centre of the sub-channel, where the measurements are being taken), so the experimentally-determined void fractions will be lower than the actual void fraction. Some of the best

  19. ABWR-II Core Design with Spectral Shift Rods for Operation with All Control Rods Withdrawn

    International Nuclear Information System (INIS)

    Moriwaki, Masanao; Aoyama, Motoo; Anegawa, Takafumi; Okada, Hiroyuki; Sakurada, Koichi; Tanabe, Akira

    2004-01-01

    An innovative reactor core concept applying spectral shift rods (SSRs) is proposed to improve the plant economy and the operability of the 1700-MW(electric) Advanced Boiling Water Reactor II (ABWR-II). The SSR is a new type of water rod in which a water level is naturally developed during operation and changed according to the coolant flow rate through the channel. By taking advantage of the large size of the ABWR-II bundle, the enhanced spectral shift operation by eight SSRs allows operation of the ABWR-II with all control rods withdrawn. In addition, the uranium-saving factor of 6 to 7% relative to the reference ABWR-II core with conventional water rods can be expected due to the greater effect of spectral shift. The combination of these advantages means the ABWR-II with SSRs should be an attractive alternative for the next-generation nuclear reactor

  20. Upon local blockage formations in LMFBR fuel rod bundles with wire-wrapped spacers

    International Nuclear Information System (INIS)

    Minden, C. v.; Schultheiss, G.F.

    1982-01-01

    A theoretical and experimental study, to improve understanding of local particle depositions in a wire-wrapped LMFBR fuel bundle, has been performed. Theoretical considerations show, that a preferentially axial process of particle depositions occurs. The experiments confirm this and clarify that the blockages arise near the particle source and settle at the spatially arranged minimum gaps in the bundle. The results suggest that, considering flow reduction, cooling and DND-detection, such fuel particle blockages are less dangerous. With reference to these safety-relevant factors, wire-wrapped LMFBR fuel bundles seem to gain advantages compared to the grid design. (orig.) [de

  1. Local hydrodynamic characteristics of regular triangular lattice of rods

    International Nuclear Information System (INIS)

    Mantlik, F.; Hejna, J.; Cervenka, J.

    1976-06-01

    Results are presented of an experimental investigation of the friction factor, velocity fields and shear stress distribution around a wetted perimeter in a rod bundle of a triangular lattice with a pitch-to-diameter ratio of 1.17. Measurements were made on 19-rod aerodynamical model at the Reynolds number of 42 300 and 211 000. The results indicated a highly significant effect of secondary flow. (author)

  2. Use of a commercial heat transfer code to predict horizontally oriented spent fuel rod surface temperatures

    International Nuclear Information System (INIS)

    Wix, S.D.; Koski, J.A.

    1993-03-01

    Radioactive spent fuel assemblies are a source of hazardous waste that will have to be dealt with in the near future. It is anticipated that the spent fuel assemblies will be transported to disposal sites in spent fuel transportation casks. In order to design a reliable and safe transportation cask, the maximum cladding temperature of the spent fuel rod arrays must be calculated. A comparison between numerical calculations using commercial thermal analysis software packages and experimental data simulating a horizontally oriented spent fuel rod array was performed. Twelve cases were analyzed using air and helium for the fill gas, with three different heat dissipation levels. The numerically predicted temperatures are higher than the experimental data for all levels of heat dissipation with air as the fill gas. The temperature differences are 4 degree C and 23 degree C for the low heat dissipation and high heat dissipation, respectively. The temperature predictions using helium as a fill gas are lower for the low and medium heat dissipation levels, but higher at the high heat dissipation. The temperature differences are 1 degree C and 6 degree C for the low and medium heat dissipation, respectively. For the high heat dissipation level, the temperature predictions are 16 degree C higher than the experimental data. Differences between the predicted and experimental temperatures can be attributed to several factors. These factors include experimental uncertainty in the temperature and heat dissipation measurements, actual convection effects not included in the model, and axial heat flow in the experimental data. This work demonstrates that horizontally oriented spent fuel rod surface temperature predictions can be made using existing commercial software packages. This work also shows that end effects will be increasingly important as the amount of dissipated heat increases

  3. Freon Rig design for performing to heat transfer experiments for nuclear reactors fuel bundles

    International Nuclear Information System (INIS)

    Flores, L.F.V.

    1981-01-01

    The main features of a Freon Rig design for performing to heat transfer experiments for PWR and BWR fuel bundles, are presented. The project is based on a Freon Rig pressurized at 30 bar with a flow rate up to 80 m 3 /h. The maximum power fed to test sections is of about 420 KW D.C. The rig was designed to use scaling techniques wich would enable a fluid of low latente heat to be used in place of water, thereby reducing the cost of testes. (Author) [pt

  4. Critical experiments supporting underwater storage of tightly packed configurations of spent fuel rods

    International Nuclear Information System (INIS)

    Hoovler, G.S.; Baldwin, M.N.

    1981-04-01

    Criticla arrays of 2.5%-enriched UO 2 fuel rods that simulate underwater rod storage of spent power reactor fuel are being constructed. Rod storage is a term used to describe a spent fuel storage concept in which the fuel bundles are disassembled and the rods are packed into specially designed cannisters. Rod storage would substantially increase the amount of fuel that could be stored in available space. These experiments are providing criticality data against which to benchmark nuclear codes used to design tightly packed rod storage racks

  5. Investigation on Minimum Film Boiling Point of Highly Heated Vertical Metal Rod in Aqueous Surfactant Solution

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chi Young; Kim, Jae Han [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2017-09-15

    In this study, experiments were conducted on the MFB(minimum film boiling) point of highly heated vertical metal rod quenched in aqueous surfactant solution at various temperature conditions. The aqueous Triton X-100 solution(100 wppm) and pure water were used as the liquid pool. Their temperatures ranged from 77 °C to 100 °C. A stainless steel vertical rod of initial center temperature of 500 °C was used as a test specimen. In both liquid pools, as the liquid temperature decreased, the time to reach the MFB point decreased with a parallel increase in the temperature and heat flux of the MFB point. However, over the whole present temperature range, in the aqueous Triton X-100 solution, the time to reach the MFB point was longer, while the temperature and heat flux of the MFB point were reduced when compared with pure water. Based on the present experimental data, this study proposed the empirical correlations to predict the MFB temperature of a high temperature vertical metal rod in pure water and in aqueous Triton X-100 solution.

  6. Experimental study of natural convective heat transfer in a vertical hexagonal sub channel

    International Nuclear Information System (INIS)

    Tandian, Nathanael P.; Umar, Efrizon; Hardianto, Toto; Febriyanto, Catur

    2012-01-01

    The development of new practices in nuclear reactor safety aspects and optimization of recent nuclear reactors, including the APWR and the PHWR reactors, needs a knowledge on natural convective heat transfer within sub-channels formed among several nuclear fuel rods or heat exchanger tubes. Unfortunately, the currently available empirical correlation equations for such heat transfer modes are limited and researches on convective heat transfer within a bundle of vertical cylinders (especially within the natural convection modes) are scarcely done. Although boundary layers around the heat exchanger cylinders or fuel rods may be dominated by their entry regions, most of available convection correlation equations are for fully developed boundary layers. Recently, an experimental study on natural convective heat transfer in a subchannel formed by several heated parallel cylinders that arranged in a hexagonal configuration has been being done. The study seeks for a new convection correlation for the natural convective heat transfer in the sub-channel formed among the hexagonal vertical cylinders. A new convective heat transfer correlation equation has been obtained from the study and compared to several similar equations in literatures.

  7. Laboratory manual for static pressure drop experiments in LMFBR wire wrapped rod bundles

    International Nuclear Information System (INIS)

    Burns, K.J.; Todreas, N.E.

    1980-07-01

    Purpose of this experiment is to determine both interior and edge subchannel axial pressure drops for a range of Reynolds numbers. The subchannel static pressure drop is used to calculate subchannel and bundle average friction factors, which can be used to verify existing friction factor correlations. The correlations for subchannel friction factors are used as input to computer codes which solve the coupled energy, continuity, and momentum equations, and are also used to develop flow split correlations which are needed as input to codes which solve only the energy equation. The bundle average friction factor is used to calculate the overall bundle pressure drop, which determines the required pumping power

  8. ORINC: a one-dimensional implicit approach to the inverse heat conduction problem. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ott, L.J.; Hedrick, R.A.

    1977-10-18

    The report develops an implicit solution technique to determine both the transient surface temperature and the transient surface heat flux of electrically heated rods given the power input and an ''indicated'' internal temperature during a simulated loss-of-coolant accident. A digital computer program ORINC (ORNL Inverse Code) is developed which solves a one-dimensional, transient, lumped parameter, implicit formulation of the conduction equation at each bundle thermocouple position in the Thermal-Hydraulic Test Facility (THTF).

  9. Implementation of a phenomenological DNB prediction model based on macroscale boiling flow processes in PWR fuel bundles

    International Nuclear Information System (INIS)

    Mohitpour, Maryam; Jahanfarnia, Gholamreza; Shams, Mehrzad

    2014-01-01

    Highlights: • A numerical framework was developed to mechanistically predict DNB in PWR bundles. • The DNB evaluation module was incorporated into the two-phase flow solver module. • Three-dimensional two-fluid model was the basis of two-phase flow solver module. • Liquid sublayer dryout model was adapted as CHF-triggering mechanism in DNB module. • Ability of DNB modeling approach was studied based on PSBT DNB tests in rod bundle. - Abstract: In this study, a numerical framework, comprising of a two-phase flow subchannel solver module and a Departure from Nucleate Boiling (DNB) evaluation module, was developed to mechanistically predict DNB in rod bundles of Pressurized Water Reactor (PWR). In this regard, the liquid sublayer dryout model was adapted as the Critical Heat Flux (CHF) triggering mechanism to reduce the dependency of the model on empirical correlations in the DNB evaluation module. To predict local flow boiling processes, a three-dimensional two-fluid formalism coupled with heat conduction was selected as the basic tool for the development of the two-phase flow subchannel analysis solver. Evaluation of the DNB modeling approach was performed against OECD/NRC NUPEC PWR Bundle tests (PSBT Benchmark) which supplied an extensive database for the development of truly mechanistic and consistent models for boiling transition and CHF. The results of the analyses demonstrated the need for additional assessment of the subcooled boiling model and the bulk condensation model implemented in the two-phase flow solver module. The proposed model slightly under-predicts the DNB power in comparison with the ones obtained from steady-state benchmark measurements. However, this prediction is acceptable compared with other codes. Another point about the DNB prediction model is that it has a conservative behavior. Examination of the axial and radial position of the first detected DNB using code-to-code comparisons on the basis of PSBT data indicated that the our

  10. Models of multi-rod code FRETA-B for transient fuel behavior analysis

    International Nuclear Information System (INIS)

    Uchida, Masaaki; Otsubo, Naoaki.

    1984-11-01

    This paper is a final report of the development of FRETA-B code, which analyzes the LWR fuel behavior during accidents, particularly the Loss-of-Coolant Accident (LOCA). The very high temperature induced by a LOCA causes oxidation of the cladding by steam and, as a combined effect with low external pressure, extensive swelling of the cladding. The latter may reach a level that the rods block the coolant channel. To analyze these phenomena, single-rod model is insufficient; FRETA-B has a capability to handle multiple fuel rods in a bundle simultaneously, including the interaction between them. In the development work, therefore, efforts were made for avoiding the excessive increase of calculation time and core memory requirement. Because of the strong dependency of the in-LOCA fuel behavior on the coolant state, FRETA-B has emphasis on heat transfer to the coolant as well as the cladding deformation. In the final version, a capability was added to analyze the fuel behavior under reflooding using empirical models. The present report describes the basic models of FRETA-B, and also gives its input manual in the appendix. (author)

  11. Large-scale transport across narrow gaps in rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Guellouz, M.S.; Tavoularis, S. [Univ. of Ottawa (Canada)

    1995-09-01

    Flow visualization and how-wire anemometry were used to investigate the velocity field in a rectangular channel containing a single cylindrical rod, which could be traversed on the centreplane to form gaps of different widths with the plane wall. The presence of large-scale, quasi-periodic structures in the vicinity of the gap has been demonstrated through flow visualization, spectral analysis and space-time correlation measurements. These structures are seen to exist even for relatively large gaps, at least up to W/D=1.350 (W is the sum of the rod diameter, D, and the gap width). The above measurements appear to compatible with the field of a street of three-dimensional, counter-rotating vortices, whose detailed structure, however, remains to be determined. The convection speed and the streamwise spacing of these vortices have been determined as functions of the gap size.

  12. CTF/STAR-CD off-line coupling for simulation of crossflow caused by mixing vane spacers in rod bundles

    International Nuclear Information System (INIS)

    Avramova, Maria

    2011-01-01

    Understanding the impact of the spacer grids on the reactor core thermal-hydraulics involves experimental mockup tests, numerical simulations, and development of reliable empirical or semi-empirical models. The state-of-the-art in modeling spacer effects on the thermal-hydraulic performance of the flow in Light Water Reactor (LWR) rod bundles employs numerical experiments by means of Computational Fluid Dynamics (CFD) calculations. The capabilities of the CFD codes are usually being validated against mock-up tests. Once validated, the CFD predictions can be used for improvement and development of more sophisticated models of the subchannel codes. Because of the involved computational cost, CFD codes can not be yet efficiently utilized for full bundle predictions, while advanced subchannel codes are a powerful tool for LWR safety and design analyses. Subchannel analyses are used for whole LWR core evaluations with relatively short CPU times and reasonable computer resources. The objectives of the presented work were to develop, implement, and qualify an innovative spacer grid model utilizing the Computational Fluid Dynamics within a framework of an efficient subchannel analysis tool. A methodology was developed for off-line coupling between the CFD code STAR-CD and the subchannel code CTF. The developed coupling scheme is flexible in axial mesh overlays. It was developed to be easily adapted to any pair of a CFD and a subchannel code. Separate modeling of the spacer grid effects on the diffusive and on the convective processes was implemented and successfully validated against experimental data. (author)

  13. Gamma densitomeric measurements of gas concentrations at a heated tube bundle; Gammadensitometrische Gasgehaltsmessungen an einem beheizten Rohrbuendel

    Energy Technology Data Exchange (ETDEWEB)

    Franz, R.; Sprewitz, U.; Hampel, U.

    2012-07-01

    The contribution under consideration reports on a gamma denitometric measurement of gas concentrations in a vertical heated tube bundle which is flowed around by a fluid. Two measurement positions, two flow rates of the circulating fluid, two subcooling values and eleven heat fluxes were selected for the measurement. The authors of this contribution describe the test facility, measurement methodology, results and their interpretation. The measurement uncertainty is described in detail.

  14. CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

    International Nuclear Information System (INIS)

    Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

    2010-01-01

    The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

  15. Experimental study on local resistance of two-phase flow through spacer grid with rod bundle

    International Nuclear Information System (INIS)

    Yan Chaoxing; Yan Changqi; Sun Licheng; Tian Qiwei

    2015-01-01

    The experimental study on local resistance of single-phase and two-phase flows through a spacer grid in a vertical channel with 3 × 3 rod bundle was carried out under the normal temperature and pressure. For the case of single-phase flow, the liquid Reynolds number covered the range of 290-18 007. For the case of two-phase flow, the ranges of gas and liquid superficial velocities were 0.013-3.763 m/s and 0.076-1.792 m/s, respectively. A correlation for predicting local resistance of single-phase flow was given based on experimental results. Eight classical two-phase viscosity formulae for homogeneous model were evaluated against the experimental data of two-phase flow. The results show that Dukler model predicts the experimental data well in the range of Re 1 < 9000 while McAdams correlation is the best one for Re 1 ≥ 9000. For all experimental data, Dukler model provides the best prediction with the mean relative error of 29.03%. A new correlation is fitted for the range of Re 1 < 9000 by considering mass quality, two- phase Reynolds number and liquid and gas densities, resulting in a good agreement with the experimental data. (authors)

  16. A comprehensive review on the methodologies to simulate the nuclear fuel bundle for the thermal hydraulic experiments

    International Nuclear Information System (INIS)

    Vishnoi, A.K.; Chandraker, D.K.; Pal, A.K.; Vijayan, P.K.; Saha, D.

    2011-01-01

    The designer of a nuclear reactor system has to ensure its safety during normal operation as well as accidental conditions. This requires, among other things, a proper understanding of the various thermal hydraulic phenomena occurring in the reactor core. In a nuclear reactor core the fuel elements are the heat source and highly loaded components of the reactor system. Therefore their behaviour under normal and accidental conditions must be extensively investigated. Data generation for Critical heat flux (CHF) in full scale bundle and parallel channel instability studies with at least two full size channels are required in order to evaluate the thermal margin and stability margin of the reactor. The complex nature of these phenomena calls for exhaustive experimental investigations. Fuel Rod Cluster Simulator (FRCS) is a very important component required for the experimental investigation of the thermal hydraulic behaviour of reactor fuel elements under normal and accidental conditions. This paper brings out a comprehensive review of the FRCS elaborating the challenges and important design aspects of the FRCS. Some of the main features and analysis results on the performance of the developed FRCS with respect to the actual nuclear fuel bundle will be presented in the paper. (author)

  17. Laboratory manual for salt-mixing test in 37- and 217-pin bundles

    International Nuclear Information System (INIS)

    Chan, Y.N.; Todreas, N.E.

    1980-08-01

    This laboratory manual deals with the procedure employed during salt tracer experiments used in evaluating the hydraulic characteristics of a rod bundle. A description of the standard equipment used is given together with the details of manufacture of probes used for detecting the salt concentration. Details of the bundle construction have been excluded as they are availble in the reference cited. An attempt has been made to point out potential trouble areas and procedures

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

    International Nuclear Information System (INIS)

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

    1984-08-01

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

  19. Study of heat transfer in 3D fuel rods of the EPRI-9R reactor modified

    International Nuclear Information System (INIS)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes

    2014-01-01

    This paper aims to conduct a case study of the fuel rods that have the highest and the lowest average power of the EPRI-9R 3D reactor modified , for various positions of the control rods banks. For this, will be addressed the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, subsequently, it is possible use the program to understand the behavior of the fuel rods and the coolant channel of the EPRI-9R 3D reactor modified. Thus, in view of the scope of this paper, first a brief introducing on the heat transfer is done, including the rod equations and the equation of energy in the channel to allow the analysis of the results

  20. Project description: ORNL PWR blowdown heat transfer separate-effects program, Thermal-Hydraulic Test Facility (THTF)

    International Nuclear Information System (INIS)

    1976-02-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results will be obtained from the Thermal-Hydraulic Test Facility (THTF), a large nonnuclear pressurized-water loop that incorporates a 49-rod electrically heated bundle. Supporting experiments will be carried out in two additional test loops - the Forced Convection Test Facility (FCTF), a small high-pressure facility in which single heater rods can be tested in annular geometry; and an air-water loop which is used to evaluate two-phase flow-measuring instrumentation

  1. Study of transient heat transfer in a fuel rod 3D, in a situation of unplanned shutdown of a PWR

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Martins, Rodolfo Ienny; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: rodolfoienny@gmail.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The study, in situations involving accidents, of heat transfer in fuel rods is of known importance, since it can be used to predict the temperature limits in designing a nuclear reactor, to assist in making more efficient fuel rods, and to increase the knowledge about the behavior of the reactor's components, a crucial aspect for safety analysis. This study was conducted using as parameter the fuel rod that has the highest average power in a typical PWR reactor. For this, we developed a program (Fuel{sub R}od{sub 3}D) in Fortran language using the Finite Elements Method (FEM) for the discretization of a fuel rod and coolant channel, in order to study the temperature distribution in both the fuel rod and the coolant channel. Transient parameters were coupled to the heat transfer equations in order to obtain details of the behavior of the rod and the channel, which allows the analysis of the temperature distribution and its change over time. This work aims to present a study case of an accident where there is a lack of energy in the reactor's coolant pumps and in the diesel engines, resulting in an unplanned shutdown of the reactor. In order to achieve the intended goal, the present work was divided as follows: a short introduction about heat transfer, including the equations concerning the fuel rod and the energy equation in the channel, an explanation about how the verification of the Fuel{sub R}od{sub 3}D program was made, and the analysis of the results. (author)

  2. Countercurrent flow-limiting characteristics of a Savannah River Plant control rod septifoil

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1992-07-01

    Experiments were performed at the Idaho National Engineering Laboratory to investigate the counter-current flow limiting characteristics of a Savannah River Plant control rod septifoil assembly. These experiments were unheated, using air and water as the working fluids. Results are presented in terms of the Wallis flooding correlation for several different control rod configurations. Flooding was observed to occur in the vicinity of the inlet slots/holes of the septifoil, rather than within the rod bundle at the location of the minimum flow area. Nearly identical flooding characteristics of the septifoil were observed for configurations with zero, three, and four rods inserted, but significantly different results occurred with 5 rods inserted

  3. Prediction of gas and liquid turbulent mixing rates between rod bundle subchannels in a two-phase slug-churn flow

    International Nuclear Information System (INIS)

    Kawahara, Akimaro; Sadatomi, Michio; Tomino, Takayoshi

    2000-01-01

    This paper presents a slug-churn flow model for predicting turbulent mixing rates of both gas and liquid phases between adjacent subchannels in a BWR fuel rod bundle. In the model, the mixing rate of the liquid phase is calculated as the sum of the three components, i.e., turbulent diffusion, convective transfer and pressure difference fluctuations between the subchannels. The components of turbulent diffusion and convective transfer are calculated from Sadatomi et al.'s (1996) method, applicable to single-phase turbulent mixing, by considering the effect of the increment of liquid velocity due to the presence of gas phase. The component of the pressure difference fluctuations is evaluated from a newly developed correlation. The mixing rate of the gas phase, on the other side, is calculated from a simple relation of mixing rate between gas and liquid phases. The validity of the proposed model has been confirmed with the turbulent mixing rates data of Rudzinski et al. as well as the present authors. (author)

  4. Effect of Tube Pitch on Pool Boiling Heat Transfer of Vertical Tube Bundle

    International Nuclear Information System (INIS)

    Kang, Myeong Gie

    2016-01-01

    Summarizing the previous results it can be stated that heat transfer coefficients are highly dependent on the tube pitch and the heat flux of the relevant tube. The published results are mostly about the horizontal tubes. However, there are many heat exchangers consisting of vertical tubes like AP600. Therefore, the focus of the present study is an identification of the effects of a tube pitch as well as the heat flux of a relevant tube on the heat transfer of a tube bundle installed vertically. When the heat flux is increased many bubbles are generating due to the increase of the nucleation sites. The bubbles become coalescing with the nearby bubbles and generates big bunches of bubbles on the tube surface. This prevents the access of the liquid to the surface and deteriorates heat transfer. The bubble coalescence is competing with the mechanisms enhancing heat transfer. The pitch was varied from 28.5 mm to 95 mm and the heat flux of the nearby tube was changed from 0 to 90kW/m"2. The enhancement of the heat transfer is clearly observed when the heat flux of the nearby tube becomes larger and the heat flux of the upper tube is less than 40kW/m"2. The effect of the tube pitch on heat transfer is negligible as the value of DP/ is increased more than 4.

  5. Axial gas transport and loss of pressure after ballooning rupture of high burn-up fuel rods subjected to LOCA conditions

    International Nuclear Information System (INIS)

    Wiesenack, Wolfgang; Oberlaender, Barbara; Kekkonen, Laura

    2008-01-01

    The OECD Halden Reactor Project has implemented integral in-pile tests on issues related to fuel behaviour under LOCA conditions. In this test series, the interaction of bonded fuel and cladding, the behaviour of fragmented fuel around the ballooning area, and the axial gas communication in high burn-up rods as affected by gap closure and fuel-clad bonding are of major interest for the investigations. In the Halden reactor tests, the decay heat is simulated by a low level of nuclear heating, in contrast to the heating conditions implemented in hot laboratory set-ups, and the thermal expansion of fuel and cladding relative to each other is more similar to the real event. The paper deals with observations regarding the loss of rod pressure following the rupture of the cladding. In the majority of the tests conducted so far, the rod pressure dropped practically instantaneously as a consequence of ballooning rupture, while one test showed a remarkably slow pressure loss. The slow loss of pressure in this test was analysed, showing that the 'hydraulic diameter' of the rod over an un-distended upper part was about 30 - 35 μm which is typical of high burn-up fuel at hot-standby conditions. The 'plug' of fuel restricts the gas flow from the plenum through the fuel column and thus limits the availability of high pressure gas for driving the ballooning. This observation is relevant for the analysis of the behaviour of a full length fuel rod under LOCA conditions since restricted gas flow may influence bundle blockage and the number of failures. (authors)

  6. Heat transfer and fluid flow in regular rod arrays with opposing flow

    International Nuclear Information System (INIS)

    Yang, J.W.

    1979-01-01

    The heat transfer and fluid flow problem of opposing flow in the fully developed laminar region has been solved analytically for regular rod arrays. The problem is governed by two parameters: the pitch-to-diameter ratio and the Grashof-to-Reynolds number ratio. The critical Gr/Re ratios for flow separation caused by the upward buoyancy force on the downward flow were evaluated for a large range of P/D ratios of the triangular array. Numerical results reveal that both the heat transfer and pressure loss are reduced by the buoyancy force. Applications to nuclear reactors are discussed

  7. PI-3 correlations and statistical evaluation results

    International Nuclear Information System (INIS)

    Pernica, R.; Cizek, J.

    1992-01-01

    Empirical Critical Heat Flux (CHF) correlations PI-3 having the widest range of validity for flow conditions in both hexagonal and square rod bundle geometries and compared with published CHF correlations are presented. They are valid for vertical water upflow through rod bundles with relatively wide and very tight rod lattices, and include axial and radial non-uniform heating. The correlations were developed with the use of more than 6000 data obtained from 119 electrically heated rod bundles. Comprehensive results of statistical evaluations of the new correlations are presented for various data bases. Also presented is a comparison of statistical evaluations of several well-known CHF correlations in the experimental data base used. A procedure which makes it possible to directly determine the probability that CHF does not occur is described for the purpose of nuclear safety assessment. (author) 8 tabs., 32 figs., 11 refs

  8. Development and adaptation of conduction and radiation heat-transfer computer codes for the CFTL

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1981-08-01

    RODCON and HOTTEL are two computational methods used to calculate thermal and radiation heat transfer for the Core Flow Test Loop (CFTL) analysis efforts. RODCON was developed at ORNL to calculate the internal temperature distribution of the fuel rod simulator (FRS) for the CFTL. RODCON solves the time-dependent heat transfer equation in two-dimensional (R angle) cylindrical coordinates at an axial plane with user-specified radial material zones and time- and position-variant surface conditions at the FRS periphery. Symmetry of the FRS periphery boundary conditions is not necessary. The governing elliptic, partial differential heat equation is cast into a fully implicit, finite-difference form by approximating the derivatives with a forward-differencing scheme with variable mesh spacing. The heat conduction path is circumferentially complete, and the potential mathematical problem at the rod center can be effectively ignored. HOTTEL is a revision of an algorithm developed by C.B. Baxi at the General Atomic Company (GAC) to be used in calculating radiation heat transfer in a rod bundle enclosed in a hexagonal duct. HOTTEL uses geometric view factors, surface emissivities, and surface areas to calculate the gray-body or composite view factors in an enclosure having multiple reflections in a nonparticipating medium

  9. Review of issues relating to enhancement of critical heat flux in nuclear reactors - 15220

    International Nuclear Information System (INIS)

    Hasan, S.M.K.; Chowdhury, M.A.Z.; Sarkar, M.A.R.

    2015-01-01

    Critical heat flux (CHF) is manifested by sharp reduction in heat transfer coefficient and a drastic rise in fuel rod surface temperature, which may lead to failure of the fuel cladding and is of immense importance in reactor design and operation. In this paper, the existing literature on CHF is studied and analyzed. Different modeling techniques such as geometry modeling and fluid modeling are highlighted. CHF enhancement techniques such as swirl flow, porous coating, nano-fluids or mechanical vibrations are analyzed and their suitability for different applications is discussed. The advances in prediction methods for tubes and rod bundles are also presented. CHF correlations and their applications have been compiled to estimate the CHF for fluids other than water or other geometries. Finally, future research needs are identified and presented. (authors)

  10. PWR-blowdown heat transfer separate effects program

    International Nuclear Information System (INIS)

    Thomas, D.G.

    1976-01-01

    The ORNL Pressurized-Water Reactor Blowdown Heat Transfer (PWR-BDHT) Program is an experimental separate-effects study of the relations among the principal variables that can alter the rate of blowdown, the presence of flow reversal and rereversal, time delay to critical heat flux, the rate at which dryout progresses, and similar time-related functions that are important to LOCA analysis. Primary test results are obtained from the Thermal-Hydraulic Test Facility (THTF). Supporting experiments are carried out in several additional test loops - the Forced Convection Test Facility (FCTF), an air-water loop, a transient steam-water loop, and a low-temperature water mockup of the THTF heater rod bundle. The studies to date are described

  11. International Standard problem ISP 14: behaviour of a fuel bundle simulator during a specified heatup and flooding period (Rebeka experiment): results of post-test analyses: final comparison report

    International Nuclear Information System (INIS)

    Karwat, H.

    1985-02-01

    The test consisted in investigating the non-steady material behaviour of a bundle of electrically heated fuel rod simulators with respect to local fuel temperatures, cladding strain, time to burst and local strain at location of burst, together with the thermal hydraulic boundary conditions. The original aim has not been fully achievable. The applied codes for mechanical fuel behaviour largely demonstrated their capabilities for pretest predictions when certain local fluid dynamic parameters are well known to the code users. The difficulties expected with proper analysis of thermal hydraulics of the test were confirmed, caused by the coupling between pin cooling conditions, rod upper plenum calculations and the feedback to clad deformation and burst simulation

  12. A model of the fluid temperature field in a turbulent flow parallel to heated tube bundle

    International Nuclear Information System (INIS)

    Carvalho Tofani, P. de.

    1986-01-01

    Basic understanding of thermal-hydraulic phenomena is essential to achieving reactor fuel assembly performance analysis. In this paper, a dimensionless parameter - a normalized fluid temperature - is defined and applied to fluid temperature measurements at particular positions at the exit plane of a bank of nine heated tubes, under different transverse heat flux shapes. This parameter presents an asymptotic trend to equilibrium values, which depend upon considered positions and flux shapes; when increasing the bulk Reynolds Number. Proposed correlations underlie the present approach to predict the fluid temperature field within the tube bundle. (Author) [pt

  13. Design Report for a 19-pin carbide test-bundle in a ring-subassembly of the test zone of KNK II/2

    International Nuclear Information System (INIS)

    Haefner, H.E.

    1982-03-01

    This report describes a 19-rod carbide test bundle in an annular oxide ring element placed at the position 201 of the test zone in the second core of KNK II as well as its behavior during the period of operation. The selected fuel rod concept includes low pellet density and a relatively large gap width as well as helium bonding between fuel and cladding. Characteristic design and operation data are: rod diameter 8.5 mm, pellet diameter 7.0 mm, maximum nominal linear rating 800 W/cm, maximum nominal burnup 70 MWd/kgHM. This report exclusively deals with the carbide test bundle and its individual components; it describes methods, criteria and results concerning the design. The annular carrier element with its head and foot is treated in a separate report. The loadability of the test bundle and its individual components is demonstrated by generally valid standards for strength criteria [de

  14. Study on heat and mass transfer characteristics of humid air-flow in a fin bundle

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong-Hwi [Air-Conditioner Research Laboratory, LG Electronics, Seoul 153-082 (Korea); Koyama, Shigeru; Kuwahara, Ken [Department of Energy and Environmental Engineering, Kyushu University, Kasuga, Fukuoka 816-8580 (Japan); Kwon, Jeong-Tae [Department of Mechanical Engineering, Hoseo University, Asan, Chungnam 336-795 (Korea); Park, Byung-Duck [School of Mechanical and Automotive Engineering, Kyungpook National University, Sangju, Gyeongbuk 742-711 (Korea)

    2010-11-15

    This paper deals with the heat and mass transfer characteristics of humid air-flow under frosting conditions. A slit fin bundle was used for the simulation of fins of a heat exchanger. The effects of the cooling block temperature, air humidity and air velocity on the frosting characteristics were experimentally investigated. The frosted mass was affected considerably by the cooling block temperature and air humidity. However, the effect of air velocity on it was not so large. The pressure drop was affected remarkably by all experimental parameters in this study. Local heat flux distribution and frost thickness distribution on each fin were predicted from the measured fin temperatures and the mass and energy conservation equations on the frost surface and inside the frost layer. (author)

  15. Thermal-hydraulic characteristics of reacting zone for TWR bundles based on CFD method

    International Nuclear Information System (INIS)

    Lu Chuan; Yan Mingyu; Lu Jianchao

    2013-01-01

    Thermal-hydraulic characteristics of reacting zone for TWR (travelling wave reactor) bundles were analysed by CFD method. The calculation results of 7, 19 and 37 fuel pin bundles show the similar characteristics. The hot coolant seems to congregate into the centre as flowing to the downstream area. The high temperature coolant always distributes in the inner area while the temperature shows distinct gradation in the outer area. The temperature difference is more than 100 ℃ for the bundle whose diameter is about 26 cm. The major temperature gradations mainly locate in the outermost fuel rods of two circles while other circles show much smaller temperature gradients. This conclusion is estimated to be true for more fuel pin bundles such as 217 fuel pin bundles. The fuel assembly structure of the existing TWR design should be optimized in future. (authors)

  16. Control rod supporting device in reactor

    International Nuclear Information System (INIS)

    Seki, Osamu; Itooka, Satoshi; Harada, Kiyoshi; Jodoi, Takashi.

    1990-01-01

    Since coolants flowing from a reactor core hit against a control rod and a control rod connection pipe, a considerable amount of bending moment for separating an attracting surface between an electromagnet and an armature is formed. Then, a plurality of grooves are formed on a heat sensitive material to dispose a heat collecting fin, and each of upper and lower contact portions of a control rod supporting portion in which the flanged portion of T-like cross section does not slip out is made into a partial spheric surface and a portion between the electromagnet and the attracted member are engaged by the unevenness. With such a constitution, even if a bending moment is applied, the control rod only swings and the bending moment is not transmitted to the attracted member. Further, since the temperature of the heat sensitive material can be rapidly made closer to the peripheral temperature by using the heat collecting fin, the timing for separation is made accurate. Further, since the engaging portion is brought into contact at the spheric surface, the load distribution on the control rod is made uniform, and the positional relationship is made accurate, to support the control rod reliably and the separation depends only on the temperature of the coolants. (N.H.)

  17. Experimental data processing technique for nonstationary heat transfer on fuel rod simulators

    International Nuclear Information System (INIS)

    Nikonov, S.P.; Nikonov, A.P.; Belyukin, V.A.

    1982-01-01

    Non-stationary heat-transfer data processing is considered in connection with experimental studies of the emergency cooling whereat fuel rod imitators both with direct and indirect shell heating were used. The objective of data processing was obtaining the temperature distribution within the imitator, the heat flux removed by the coolant and the shell-coolant heat-transfer coefficient. The special attention was paid to the temperature distribution calculation at the data processing during the reflooding experiments. In this case two factors are assumed to be known: the time dependency of temperature variation at a certain point within the imitator cross-section and the heat flux at some point of the same cross-section. The initial data preparation for calculations, employing the procedure of smoothing by cubic spline functions, is considered as well, with application of an algorithm reported in the literature, which is efficient for the given functional dependency wherein the deviation in each point is known [ru

  18. Fluid-mixing studies in a hexagonal 61-pin wire-wrapped rod bundle

    International Nuclear Information System (INIS)

    Symolon, P.D.; Todreas, N.E.

    1981-02-01

    Mixing, pressure drop, and flow split experiments were performed on a 61 pin LMFBR fuel bundle with a pitch to diameter ratio of 1.25, and lead lengths of 6 and 12 inches. The mixing results obtained from salt injection experiments were found to depend on injection depth. The deeper injection is expected to give the more accurate results. The pressure drop data was presented as friction factor versus Reynolds number, and the results were compared to the correlation of Hawley. The flowsplit data presented was flawed by corroded bundle walls, but some insight was obtained on the effect of rough surfaces on flowsplit, and how to account for its effect in the correlations

  19. Protective guide structure for reactor control rod

    International Nuclear Information System (INIS)

    Ban, Minoru; Umeda, Kenji; Kubo, Noboru; Ito, Tomohiro.

    1996-01-01

    The present invention provides an improved protective guide structure for control rods, which does not cause swirling of coolants and resonance even though a slit is formed on a protective tube which surrounds a control rod element in a PWR type reactor. Namely, a reactor control rod is constituted with elongated control elements collectively bundled in the form of a cluster. The protective guide structure protectively guides the collected constituent at the upper portion of a reactor container. The protective structure comprises a plurality of protective tubes each having a C-shaped cross section disposed in parallel for receiving control rod elements individually in which the corners of the opening of the cross section of the protective tube are chamfered to an appropriate configuration. With such a constitution, even if coolant flows in a circumferential direction along the protective tubes surrounding the control rod elements, no shearing stream is caused to the coolants flow since the corners of the cross sectional opening (slit) of the tube are chamfered. Accordingly, occurrence of swirlings can be suppressed. (I.S.)

  20. Development of a new bundle welding technology for CANDU fuels

    International Nuclear Information System (INIS)

    Kim, Soo Sung; Lee, D. Y.; Goo, D. S.

    2010-01-01

    The new technology of welding process for fuel bundle of CANDU nuclear fuels is considered important in respect to the soundness of weldments and the improvement of the performance of nuclear fuels during the operation in reactor. The probability of leakage of the fission products is mostly apt to occur at the weldments of fuel bundles, and it is connected directly with the safety and life prediction of the nuclear reactor in operation. The fuel bundles of CANDU nuclear fuels are welded by the electrical resistance method, connecting the endplates and endcaps with fuel rods. Therefore, the purpose of this study of the 2nd year is to select the proper welding parameters and to investigate the characteristics of the full-sized samples using the projection endplates and make some prototype samples for the endplate welding of CANDU nuclear fuels. This study will be also provide the fundamental data for the new design and fabrications of CANDU nuclear fuel bundles