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Sample records for lwr rod bundles

  1. 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

  2. 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)

  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. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    International Nuclear Information System (INIS)

    Sample, C.R.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL

  5. Literature search on Light Water Reactor (LWR) fuel and absorber rod fabrication, 1960--1976

    Energy Technology Data Exchange (ETDEWEB)

    Sample, C R [comp.

    1977-02-01

    A literature search was conducted to provide information supporting the design of a conceptual Light Water Reactor (LWR) Fuel Fabrication plant. Emphasis was placed on fuel processing and pin bundle fabrication, effects of fuel impurities and microstructure on performance and densification, quality assurance, absorber and poison rod fabrication, and fuel pin welding. All data have been taken from publicly available documents, journals, and books. This work was sponsored by the Finishing Processes-Mixed Oxide (MOX) Fuel Fabrication Studies program at HEDL.

  6. 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

  7. 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

  8. 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)

  9. 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

  10. Evaluation of LWR fuel rod behavior under operational transient conditions

    International Nuclear Information System (INIS)

    Nakamura, M.; Hiramoto, K.; Maru, A.

    1984-01-01

    To evaluate the effects of fission gas flow and diffusion in the fuel-cladding gap on fuel rod thermal and mechanical behaviors in light water reactor (LWR) fuel rods under operational transient conditions, computer sub-programs which can calculate the gas flow and diffusion have been developed and integrated into the LWR fuel rod performance code BEAF. This integrated code also calculates transient temperature distribution in the fuel-pellet and cladding. The integrated code was applied to an analysis of Inter Ramp Project data, which showed that by taking into account the gas flow and diffusion effects, the calculated cladding damage indices predicted for the failed rods in the ramp test were consistent with iodine-SCC (Stress Corrosion Cracking) failure conditions which were obtained from out-of-reactor pressurized tube experiments with irradiated Zircaloy claddings. This consistency was not seen if the gas flow and diffusion effects were neglected. Evaluation were also made for the BWR 8x8 RJ fuel rod temperatures under power ramp conditions. (orig.)

  11. 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.

  12. 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

  13. Core design of super LWR with double tube water rods

    International Nuclear Information System (INIS)

    Wu, Jianhui; Oka, Yoshiaki

    2014-01-01

    Highlights: • Supercritical light water cooled and moderated reactor with double tube water rods is developed. • Double-row fuel rod assembly and out-in fuel loading pattern are applied. • Separation plates in peripheral assemblies increase average outlet temperature. • Neutronic and thermal design criteria are satisfied during the cycle. - Abstract: Double tube water rods are employed in core design of super LWR to simplify the upper core structure and refueling procedure. The light water moderator flows up in the inner tube from the bottom of the core, then, changes the flow direction at the top of the core into the outer tube and flows out at the bottom of the core. It eliminates the moderator guide/distribution tubes into the single tube water rods from the top dome of the reactor pressure vessel of the previous super LWR design. Two rows of fuel rods are filled between the water rods in the fuel assembly. Out-in refueling pattern is adopted to flatten radial power distribution. The peripheral fuel assemblies of the core are divided into four flow zones by separation plates for increasing the average core outlet temperature. Three enrichment zones are used for axial power flattening. The equilibrium core is analyzed based on neutronic/thermal-hydraulic coupled model. The results show that, by applying the separation plates in peripheral fuel assemblies and low gadolinia enrichment, the maximum cladding surface temperature (MCST) is limited to 653 °C with the average outlet temperature of 500 °C. The inherent safety is satisfied by the negative void reactivity effects and sufficient shutdown margin

  14. 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

  15. 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.

  16. 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.)

  17. 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

  18. 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

  19. 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)

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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)

  5. 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)

  6. 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.

  7. 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

  8. 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

  9. 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)

  10. 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

  11. 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.

  12. 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

  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. Performance of artificially defected LWR fuel rods in an unlimited air dry storage atmosphere

    International Nuclear Information System (INIS)

    Einziger, R.E.; Knecht, R.L.; Cantley, D.A.; Cook, J.A.

    1983-09-01

    Thus far the tests are inconclusive as to whether breached LWR fuel can be stored at 230 0 C for long periods of time in air without fuel oxidation and dispersion. There is every indication, as expected, that there is no oxidation problem in an inert atmosphere. Only one of four defects exposed to unlimited air gave any indication of fuel oxidation. It has been suggested that this might be an incubation effect and continued operation would result in oxidation occurring at all four defects. As yet the destructive examination of the BWR rod has not been completed, so it is not possible to determine if cladding splitting was due to an anomoly in this test rod or something that can be expected in LWR rods in general. Thus far there is no indication of respirable particle dispersal even if fuel oxidation does occur

  15. 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)

  16. Measurement of blockage in deformed LWR multi-rod arrays

    International Nuclear Information System (INIS)

    Hindle, E.D.; Jones, C.; Whitty, S.

    1983-01-01

    This paper critically reviews the current methods used for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed. Also examples of the application of automatic computerised techniques to directly measure rod strain, blockage, sub-channel blockage and perimeter changes from photographs of sections through deformed arrays are presented. (author)

  17. 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

  18. 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

  19. 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.

  20. 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

  1. Axial transport of fission gas in LWR fuel rods

    International Nuclear Information System (INIS)

    Kinoshita, M.

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  3. Gap conductance in Zircaloy-clad LWR fuel rods

    International Nuclear Information System (INIS)

    Ainscough, J.B.

    1982-04-01

    This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO 2 in Zircaloy cladding, under both steady-state and transient conditions. The relevant theory is discussed together with some of the approximations usually made in performance modelling codes. The state of the physical property data which are needed for heat transfer calculations is examined and some of the relevant in- and out-of-reactor experimental work on fuel rod conductance is reviewed

  4. 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)

  5. 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)

  6. 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

  7. 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)

  8. 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

  9. 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

  10. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Purcell, P.C. [BNFL International Transport, Spent Fuel Services (United Kingdom); Dallongeville, M. [COGEMA Logistics (AREVA Group) (France)

    2004-07-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme.

  11. Testing of LWR fuel rods to support criticality safety analysis of transport accident conditions

    International Nuclear Information System (INIS)

    Purcell, P.C.; Dallongeville, M.

    2004-01-01

    For the transport of low enriched materials, criticality safety may be demonstrated by applying pessimistic modelling assumptions that bound any realistic case. Where Light Water Reactor (LWR) fuel is being transported, enrichment levels are usually too high to permit this approach and more realistic data is needed. This requires a method by which the response of LWR fuel under impact accident conditions can be approximated or bounded. In 2000, BNFL and COGEMA LOGISTICS jointly commenced the Fuel Integrity Project (FIP) whose objective was to develop such methods. COGEMA LOGISTICS were well advanced with a method for determining the impact response of unirradiated fuel, but required further test data before acceptance by the Transport Regulators. The joint project team extensively discussed the required inputs to the FIP, from which it was agreed that BNFL would organise new tests on both unirradiated and irradiated fuel samples and COGEMA LOGISTICS would take major responsibility for evaluating the test results. Tests on unirradiated fuel rod samples involved both dynamic and quasi-static loading on fuel samples. PWR fuel rods loaded with uranium pellets were dropped vertically from 9m onto a rigid target and this was repeated on BWR fuel rods, similar tests on empty fuel rods were also conducted. Quasi-static tests were conducted on 530 mm long PWR and BWR fuel specimens under axial loading. Tests on irradiated fuel samples were conducted on high burn-up fuel rods of both PWR and BWR types. These were believed original to the FIP project and involved applying bending loads to simply supported pressurised rod specimens. In one test the fuel rod was heated to nearly 500oC during loading, all specimens were subject to axial impact before testing. Considerable experience of fuel rod testing and new data was gained from this test programme

  12. 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)

  13. Evaluation of fuel rod damage in LWR under accident conditions using SSYST

    International Nuclear Information System (INIS)

    Meyder, R.

    1982-01-01

    After a short outline of the recent SSYST-development, the creep rupture model NORA 2 is presented. The effect of temperature and oxygen on Zircaloy 4 creep behaviour is shown. Examples on the effect of azimuthal varying gap width and wall thickness are given. Remarks on the extension of a single rod analysis on a bundle and the stepwise application of SSYST for investigation of fuel rod failure conclude the paper. (orig.) [de

  14. 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)

  15. Estimation of the core-wide fuel rod damage during a LWR LOCA

    International Nuclear Information System (INIS)

    Mattila, L.; Sairanen, R.; Stengaard, J.-O.

    1975-01-01

    The number of fuel rods puncturing during a LWR LOCA must be estimated as a part of the plant radioactivity release analysis. Due to the great number of fuel rods in the core and the great number of contributing parameters, many of them associated with wide uncertainty and/or truly random variability limits, probabilistic methods are well applicable. A succession of computer models developed for this purpose is described together with applications to WWER-440 PWR. Deterministic models are shown to be seriously inadequate and even misleading under certain circumstances. A simple analytical probabilistic model appears to be suitable for many applications. Monte Carlo techniques allow the development of such sophisticated models that errors in the input data presently available probably become dominant in the residual uncertainty of the corewide fuel rod puncture analysis. (author)

  16. 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)

  17. 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

  18. Experimental critical loadings and control rod worths in LWR-PROTEUS configurations compared with MCNPX results

    International Nuclear Information System (INIS)

    Plaschy, M.; Murphy, M.; Jatuff, F.; Seiler, R.; Chawla, R.

    2006-01-01

    The PROTEUS research reactor at the Paul Scherrer Inst. (PSI) has been operating since the sixties and has already permitted, due to its high flexibility, investigation of a large range of very different nuclear systems. Currently, the ongoing experimental programme is called LWR-PROTEUS. This programme was started in 1997 and concerns large-scale investigations of advanced light water reactors (LWR) fuels. Until now, the different LWR-PROTEUS phases have permitted to study more than fifteen different configurations, each of them having to be demonstrated to be operationally safe, in particular, for the Swiss safety authorities. In this context, recent developments of the PSI computer capabilities have made possible the use of full-scale SD-heterogeneous MCNPX models to calculate accurately different safety related parameters (e.g. the critical driver loading and the shutdown rod worth). The current paper presents the MCNPX predictions of these operational characteristics for seven different LWR-PROTEUS configurations using a large number of nuclear data libraries. More specifically, this significant benchmarking exercise is based on the ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JENDL3.2, and JENDL3.3 libraries. The results highlight certain library specific trends in the prediction of the multiplication factor k eff (e.g. the systematically larger reactivity calculated with JEF2.2 and the smaller reactivity associated with JEFF3.0). They also confirm the satisfactory determination of reactivity variations by all calculational schemes, for instance, due to the introduction of a safety rod pair, these calculations having been compared with experiments. (authors)

  19. 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)

  20. 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.

  1. Characterization of LWR fuel rod irradiations with power transients in the BR2 reflector

    International Nuclear Information System (INIS)

    Ponsard, B.; Bodart, S.; Meer, K. van der; Raedt, C. de

    1996-01-01

    Fuel rod irradiations in reflector positions of the materials testing reactor BR2 are becoming increasingly important. A typical example is that of irradiation devices containing single LWR fuel rods, to be tested in the framework of a new international fuel investigation and development programme. Some of the irradiations will comprise power transients with central fuel melting (at 2800 deg. C), the power increase being obtained by decreasing the pressure in a He-3 neutron absorbing screen and/or by varying the BR2 reactor operating power. A total power variation by a factor of at least 2.5 in the fuel rod irradiated could thus be achieved. In some of the rods, central temperature measurements (up to 2000 deg. C) will be carried out. Both fresh and pre-irradiated fuel rods are concerned in the programme. For these irradiations, the accurate knowledge of the neutron-induced fission heating and of the gamma heating is required, as one of the purposes of the programme consists in establishing the correlation among the thermal conductivity, the burn-up and the irradiation temperature. Calibration work among various measuring methods and between measurements and one- and two-dimensional calculations is being pursued. (author). 10 refs, 15 figs, 3 tabs

  2. 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

  3. 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

  4. 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.

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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.)

  10. 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

  11. Three dimensional considerations in thermal-hydraulics of helical cruciform fuel rods for LWR power uprates

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush, E-mail: kshirvan@mit.edu; Kazimi, Mujid S.

    2014-04-01

    Highlights: • We benchmarked the 4 × 4 helical cruciform fuel (HCF) bundle pressure drop experimental data with CFD. • We also benchmarked the 4 × 4 HCF mixing experimental data with CFD. • We derived new friction factors for PWR and BWR designs at PWR and BWR operating conditions from CFD. • We showed the importance of modeling the 3D conduction in HCF in steady state and transient conditions. - Abstract: In order to increase the power density of current and new light water reactor designs, the helical cruciform fuel (HCF) rods have been proposed. The HCF rod is equivalent to a thin cylindrical rod, with 4 fuel containing vanes, wrapped around it. The HCF rods increase the surface area to volume ratio of the fuel and enhance the inter-subchannel mixing due to their helical shape. The rods do not need supporting grids, as they are packed to periodically contact their neighbors along the flow direction, enabling a higher power density in the core. The HCF rods were reported to have the potential to uprate existing PWRs by 45% and BWRs by 20%. In order to quantify the mixing behavior of the HCF rods based on their twist pitch, experiments were previously performed at atmospheric pressures with single phase water in a 4 by 4 HCF and cylindrical rod bundles. In this paper, the experimental results on pressure drop and mixing are benchmarked with computational fluid dynamic (CFD) using steady state the Reynolds average Navier–Stokes (RANS) turbulence model. The sensitivity of the CFD approach to computational domain, mesh size, mesh shape and RANS turbulence models are examined against the experimental conditions. Due to the refined radial velocity profile from the HCF rods twist, the turbulence models showed little sensitivity to the domain. Based on the CFD simulations, the total pressure drops under the PWR and BWR conditions are expected to be about 10% higher than the values previously reported solely from an empirical correlation based on the

  12. 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)

  13. 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

  14. 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)

  15. 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)

  16. 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

  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. 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

  19. SSYST. A code system to analyze LWR fuel rod behavior under accident conditions

    International Nuclear Information System (INIS)

    Gulden, W.; Meyder, R.; Borgwaldt, H.

    1982-01-01

    SSYST (Safety SYSTem) is a modular system to analyze the behavior of light water reactor fuel rods and fuel rod simulators under accident conditions. It has been developed in close cooperation between Kernforschungszentrum Karlsruhe (KfK) and the Institut fuer Kerntechnik und Energiewandlung (IKE), University Stuttgart, under contract of Projekt Nukleare Sicherheit (PNS) at KfK. Although originally aimed at single rod analysis, features are available to calculate effects such as blockage ratios of bundles and wholes cores. A number of inpile and out-of-pile experiments were used to assess the system. Main differences versus codes like FRAP-T with similar applications are (1) an open-ended modular code organisation, (2) availability of modules of different sophistication levels for the same physical processes, and (3) a preference for simple models, wherever possible. The first feature makes SSYST a very flexible tool, easily adapted to changing requirements; the second enables the user to select computational models adequate to the significance of the physical process. This leads together with the third feature to short execution times. The analysis of transient rod behavior under LOCA boundary conditions e.g. takes 2 mins cpu-time (IBM-3033), so that extensive parametric studies become possible

  20. 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)

  1. LWR fuel rod testing facilities in high flux reactor (HFT) Petten for investigation of power cycling and ramping behaviour

    Energy Technology Data Exchange (ETDEWEB)

    Markgraf, J; Perry, D; Oudaert, J [Commission of the European Communities, Joint Reserach Centre, Petten Establishment, Petten (Netherlands)

    1983-06-01

    LWR fuel rod irradiation experiments are being performed in HFR's Pool Side Facility (PSF) by means of pressurized boiling water capsules (BWFC). For more than 6 years the major application of these devices has been in performing irradiation programs to investigate the power ramp behaviour of PWR and BWR rods which have been pre-irradiated in power reactors. Irradiation devices with various types of monitoring instrumentation are available, e.g. for fuel rod length, fuel stack length, fuel rod internal pressure and fuel rod central temperature measurements. The application scope covers PWR and BWR fuel rods, with burn-up levels up to 45 MWd/kg(U), max. linear heat generation rates up to 700 W/cm and simulation of constant power change rates between 0.05 and 1000 W/cm.min. The paper describes the different designs of LWR fuel rod testing facilities and associated non-destructive testing devices in use at the HFR Petten. It also addresses the new test capabilities that will become available after exchange of the HFR vessel in 1983. Furthermore it shows some typical results. (author)

  2. 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.

  3. 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.

  4. 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

  5. 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

  6. 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)

  7. 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)

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

    International Nuclear Information System (INIS)

    Shen Xiuzhong; Yu Pingan; Yang Guanyue

    1996-01-01

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

  9. 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

  10. 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)

  11. 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)

  12. 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

  13. 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.

  14. PLUTON, Isotope Generation and Depletion in Highly Irradiated LWR Fuel Rods

    International Nuclear Information System (INIS)

    Lemehov, Sergei; Motoe, Suzuki

    2003-01-01

    1 - Description of program or function: The PLUTON-PC is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO 2 , UO 2 -Gd 2 O 3 , inhomogeneous MOX, and UO 2 -ThO 2 . The PLUTON-PC code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. 2 - Methods: Based upon cumulative yields, the PLUTON-PC code calculates as a function of radial position and local burnup concentrations of fission products, macroscopic scattering cross-sections and self-shielding effect which is important for standard fuel (for Pu-242 mainly) and more importantly for homogeneous and inhomogeneous MOX fuel because of higher concentrations of fissile and fertile isotopes of plutonium. The code results in burnup dependent fission rate density profiles throughout the in-reactor irradiation of LWR fuel rods. The isotopes included in calculations have been extended to cover all trans-uranium groups (plutonium plus higher actinides) of fissile and fertile isotopes. Self-shielding problem and scattering effects have been revised and solved for all isotopes in the calculations for adequacy at high burnup, different irradiation conditions and cladding materials

  15. 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)

  16. 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

  17. 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

  18. 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

  19. 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

  20. Characterisation of high-burnup LWR fuel rods through gamma tomography

    International Nuclear Information System (INIS)

    Caruso, S.

    2007-01-01

    Current fuel management strategies for light water reactors (LWRs), in countries with high back-end costs, progressively extend the discharge burnup at the expense of increasing the 235 U enrichment of the fresh UO 2 fuel loaded. In this perspective, standard non-destructive assay techniques, which are very attractive because they are fast, cheap, and preserve the fuel integrity, in contrast to destructive approaches, require further validation when burnup values become higher than 50 GWd/t. This doctoral work has been devoted to the development and optimisation of non-destructive assay techniques based on gamma-ray emissions from irradiated fuel. It represents an important extension of the unique, high-burnup related database, generated in the framework of the LWR PROTEUS Phase II experiments. A novel tomographic measurement station has been designed and developed for the investigation of irradiated fuel rod segments. A unique feature of the station is that it allows both gamma-ray transmission and emission computerised tomography to be performed on single fuel rods. Four burnt UO 2 fuel rod segments of 400 mm length have been investigated, two with very high (52 GWd/t and 71 GWd/t) and two with ultra-high (91 GWd/t and 126 GWd/t) burnup. Several research areas have been addressed, as described below. The application of transmission tomography to spent fuel rods has been a major task, because of difficulties of implementation and the uniqueness of the experiments. The main achievements, in this context, have been the determination of fuel rod average material density (a linear relationship between density and burnup was established), fuel rod linear attenuation coefficient distribution (for use in emission tomography), and fuel rod material density distribution. The non-destructive technique of emission computerised tomography (CT) has been applied to the very high and ultra-high burnup fuel rod samples for determining their within-rod distributions of caesium and

  1. 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)

  2. 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

  3. 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)

  4. 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)

  5. 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.

  6. 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.)

  7. 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.

  8. 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.

  9. 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)

  10. 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)

  11. 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.

  12. 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.)

  13. Status and results of the theoretical and experimental investigations on the LWR fuel rod behavior under accident conditions

    International Nuclear Information System (INIS)

    Bocek, M.; Hofmann, P.; Leistikow, S.; Class, G.; Meyder, R.; Raff, S.; Erbacher, F.; Hofmann, G.; Ihle, P.; Karb, E.; Fiege, A.

    1978-09-01

    In this report the status of knowledge is described which has been gathered up to the end of 1977 of the LWR fuel rod behavior in loss-of-coolant accidents. The majority of results indicated have been derived from studies on the fuel rod behavior performed within the framework of the Nuclear Safety Project (PNS); partly, also the results of cooperating research establishments and fm international exchange of experience are referred to. The report has been subdivided into two complete parts: Part I provides a survey of the most significant results of the theoretical and experimental research projects on fuel rod behavior. Part II describes by detailed individual presentations the status as well as the results with respect to the major central subjects. (orig.) 891 RW 892 AP [de

  14. 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.

  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. 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

  17. 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

  18. 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.

  19. 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.

  20. 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)

  1. Influence of some fabrication parameters and operating conditions on the PCI failure occurrence in LWR fuel rods

    International Nuclear Information System (INIS)

    Bouffioux, P.

    1980-01-01

    In recent LWR designs, the fuel rod failures are induced by a chemically assisted mechanical process, i.e. stress corrosion cracking. The analytical approach towards the analysis of PCI-SCC failures is mainly based on the predictions of the COMETHE code. The failure criteria rely on the concept of a stress threshold together with fission product availability. In the present paper, the use of the COMETHE code to minimize PCI induced clad failure occurrences is illustrated by parametric studies to define acceptable fuel specifications and reactor operating conditions (steady and transient). (author)

  2. RADGEN: A radiation exchange factor generator for rod bundles

    International Nuclear Information System (INIS)

    Rector, D.R.

    1987-10-01

    The RADGEN computer program has been developed at Pacific Northwest Laboratory (PNL) to generate input required for the thermal radiation models used in the COBRA-SFS (Spent Fuel Storage) computer program. The COBRA-SFS program uses radiation exchange factors to describe the net amount of energy transferred from each surface to every other surface in an enclosure. The RADGEN program generates radiation exchange factors for arrays of rods on a square or triangular pitch as well as open channel geometries. This report describes the input requirements for the RADGEN code, which may be executed in a batch or interactive mode, and outlines the solution procedure used to obtain the exchange factors. 4 refs., 25 figs., 13 tabs

  3. 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)

  4. 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)

  5. 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)

  6. 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

  7. 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)

  8. Single-Phase Crossflow Mixing in a Vertical Tube Bundle Geometry : An Experimental Study

    NARCIS (Netherlands)

    Mahmood, A.

    2011-01-01

    The vertical rod/tube bundle geometry has a wide variety of industrial applications. Typical examples are the core of light water nuclear reactors (LWR) and vertical tube steam generators. In the core of a LWR, primarily coolant flows upward but their also exist a flow in lateral direction, called

  9. 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)

  10. 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.)

  11. 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)

  12. 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.

  13. 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.

  14. 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

  15. 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

  16. 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.)

  17. 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

  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. 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)

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. 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)

  5. 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)

  6. 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.

  7. 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)

  8. 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

  9. 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

  10. Results from In-pile experiments on LWR fuel rod behavior under LOCA conditions with unirradiated rods

    International Nuclear Information System (INIS)

    Sepold, L.; Karb, E.H.; Pruessmann, M.

    1981-06-01

    This report summarizes the results of the FR2-in-pile tests at KfK (Kernforschungszentrum Karlsruhe) with unirradiated test rods. The in-pile tests with the objective of investigating the influence of a nuclear environment on the mechanisms of fuel rod failure were being performed with irradiated and unirradiated single rods of a PWR design in the DK loop of the FR2 reactor. The main parameter of the test program was the burnup, ranging from 2.500 to 35.000 MWd/t. The program with unirradiated specimens comprised the series A and B with a total of 14 tests. (orig.) [de

  11. Behavior of defective LWR-type fuel rods irradiated under postulated accident conditions

    International Nuclear Information System (INIS)

    Hobbins, R.R.; Croucher, D.W.; Seiffert, S.L.; Cook, B.A.; Kerwin, D.K.; Mehner, A.S.; Ploger, S.A.

    1979-05-01

    The irradiation experiments reported here have been conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc., for the United States Nuclear Regulatory Commission in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory. Five of the rods were irradiated in PCM tests and one in a LOC test. During these tests, the six rods lost cladding integrity prior to or during the transient phase of the test due to either manufacturing defects or intentional rod design and operation. Of the five defective rods tested under PCM conditions, one (Rod IE-008, Test IE-1) had a hydride rupture below the region of the rod, which was in film boiling during the transient; two (Rod A-0021, Test PCM-3 and Rod IE-019, Test IE-5) contained defects (a pin hole and a small axial crack, respectively) within the film boiling zone; and two (Rod 201-1, Test PCM-1 and Rod 205-8, Test PCM-5) failed by cladding embrittlement within the film boiling zone. Rod 312-3 was waterlogged before being subjected to LOC conditions in Test LLR-3

  12. 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

  13. 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

  14. 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)

  15. Theoretical investigations of the gas flow in ballooning LWR-fuel rods

    International Nuclear Information System (INIS)

    Gaballah, I.

    1978-09-01

    A theory is developed for the calculation of gas flow in a fuel rod simulator or in a fuel rod with round- or cracked pellets. The fundamental equations are formulated, simplified, reformed, and then numerically solved. The numerical investigations show, that a quasi steady incompressible flow model can be used without great error. The effect of the deformation form is studied. A uniform deformation along the whole length causes small pressure difference. A power profile and rod spacers cause non-uniform clad deformation of the fuel rod simulator or the fuel rod. This deformation leads to greater pressure differences. Finally the effect of the cracked pellets is studied. The cracked pellets cause great pressure differences along the fuel rod. (orig.) 891 HP [de

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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.

  6. Influence of pellet-clad-gap-size on LWR fuel rod performance

    International Nuclear Information System (INIS)

    Brzoska, B.; Fuchs, H.P.; Garzarolli, F.; Manzel, R.

    1979-01-01

    The as-fabricated pellet-clad-gap size varies due to fabricational tolerances of the cladding inner diameter and the pellet outer diameter. The consequences of these variations on the fuel rod behaviour are analyzed using the KWU fuel rod code CARO. The code predictions are compared with experimental results of special pathfinder test fuel rods irradiated in the OBRIGHEIM nuclear power plant. These test fuel rods include gap sizer in the range of 140 μm to 270 μm, prepressurization between 13 bar to 36 bar and Helium and Argon fill gases irradiated up to a local burnup of 35 MWd/kg(U). Post irradiation examination were performed at different burnups. CARC calculations have been performed with special emphasis in cladding creep down, fission gas release and pellet clad gap closure. (orig.)

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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)

  19. 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)

  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. 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

  2. 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)

  3. 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

  4. The ''THERMOST'' for analysing thermo-structural behaviour of LWR fuel rod under PCI conditions

    International Nuclear Information System (INIS)

    Nuno, H.; Ogawa, S.; Kobayashi, H.

    1983-01-01

    As one of the methods for evaluating the fuel rod performances under power ramping or load following operations, the combined ''FROST'' and ''THERMOST'' system has been developed and being brought into practical use. The former had already been presented at Blackpool Meeting in 1978, and the latter is going to be presented in this paper. The major purpose of the THERMOST is to analyse very detailed thermal and structural fuel behaviours in a rather localized part of fuel rod whereas the FROST deals with whole-rod-wide general performances. The code handles 2-dimensional thermal and structural analyses simultaneously by using finite element method, in axial section wide or in lateral section wide. It consists of a fundamental FEM system of generalized constitution and its surrounding subroutine system which characterizes fuel behaviours such as temperature distribution, thermal expansion, elastoplasticity, creep, cracking, swelling, growth, etc. Thermal analysis is handled by heat conduction and heat transfer elements (6 kinds) and structural analysis by axisymmetric ring and lateral plane elements (6 kinds). Boundary problems such as contact, friction and cracking are treated by gap and crack elements. A sample calculation of PCI performance on a PWR fuel rod under ramping condition is presented with some inpile test data. (author)

  5. 'THERMOST' for analysing thermo-structural behaviour of LWR fuel rods under PCI conditions

    International Nuclear Information System (INIS)

    Nuno, H.; Ogawa, S.; Kobayashi, H.

    1983-01-01

    As a method for evaluating fuel rod performance under power ramping or load following operations, the combined FROST/ THERMOST system has been developed and brought into practical use. FROST was presented at the IAEA Blackpool Meeting in 1978, and THERMOST is the subject of this paper. The major purpose of THERMOST is to analyse very detailed thermal and structural fuel behaviour in a rather localised part of the fuel rod whereas FROST deals with whole rod general performance. The code handles two-dimensional thermal and structural analyses simultaneously by using a finite element method, in axial section or in lateral section. It consists of a fundamental FEM system of generalised constitution, and a surrounding subroutine system which characterises fuel behaviour, such as temperature distribution, thermal expansion, elastoplasticity, creep, cracking, swelling, growth, etc. Thermal analysis is handled by heat conduction and heat transfer element (six kinds), and structural analysis by axisymmetric ring and lateral plane element (six kinds). Boundary problems such as contact, friction and cracking are treated by gap and crack elements. A sample calculation of PCI performance on a PWR fuel rod under ramping conditions is presented with some in-pile test data. (author)

  6. Fission Gas Release in LWR Fuel Rods Exhibiting Very High Burn-Up

    DEFF Research Database (Denmark)

    Carlsen, H.

    1980-01-01

    Two UO2Zr BWR type test fuel rods were irradiated to a burn-up of about 38000 MWd/tUO2. After non-destructive characterization, the fission gas released to the internal free volume was extracted and analysed. The irradiation was simulated by means of the Danish fuel performance code WAFER-2, which...

  7. 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

  8. 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

  9. Thermal performance of annular-coated and sphere-pac LWR fuel rod designs

    International Nuclear Information System (INIS)

    Guenther, R.J.; Hsieh, K.A.; Barner, J.O.; Freshley, M.D.

    1980-01-01

    Two FCI-resistant UO 2 fuel rod designs are being compared to a reference design in irradiation tests in the Halden Boiling Water Reactor (HBWR) as part of the DOE-sponsored Fuel Performance Improvement Program (FPIP). The primary fuel design (annular-coated-pressurized) incorporates annular pellets, a graphite coating on the inner surface of the Zircaloy cladding, and pressurized helium fill gas. Also being investigated is an 87% smear density sphere-pac design with pressurized helium fill gas. The solid pellet (reference) and annular-coated designs described had helium fill gas at approx. 100 kPa and the sphere-pac rods were pressurized at approx. 455 kPa

  10. Cladding temperature measurement by thermocouples at preirradiated LWR fuel rod samples

    International Nuclear Information System (INIS)

    Leiling, W.

    1981-12-01

    This report describes the technique to measure cladding temperatures of test fuel rod samples, applied during the in-pile tests on fuel rod failure in the steam loop of the FR2 reactor. NiCr/Ni thermocouples with stainless steel and Inconel sheaths, respectively,of 1 mm diameter were resistance spot weld to the outside of the fuel rod cladding. For the pre-irradiated test specimens, welding had to be done under hot-cell conditions, i.e. under remote handling. In order to prevent the formation of eutectics between zirconium and the chemical elements of the thermocouple sheath at elevated temperatures, the thermocouples were covered with a platinum jacket of 1.4 mm outside diameter swaged onto the sheath in the area of the measuring junction. This thermocouple design has worked satisfactorily in the in-pile experiments performed in a steam atmosphere. Even in the heatup phase, in which cladding temperatures up to 1050 0 C were reached, only very few failures occured. This good performance is to a great part due to a careful control and a thorough inspection of the thermocouples. (orig.) [de

  11. A study on gap heat transfer of LWR fuel rods under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    Fujishiro, Toshio

    1984-03-01

    Gap heat transfer between fuel pellet and cladding have a large influence on the LWR fuel behaviors under reactivity initiated accident (RIA) conditions. The objective of the present study is to investigate the effects of gap heat transfer on RIA fuel behaviors based on the results of the gap-gas parameter tests in NSRR and on their analysis with NSR-77 code. Through this study, transient variations of gap heat transfer, the effects of the gap heat transfer on fuel thermal behaviors and on fuel failure, effects of pellet-cladding sticking by eutectic formation, and the effects of cladding collapse under high external pressure have been clearified. The studies have also been performed on the applicability and its limit of modified Ross and Stoute equation which is extensively utilized to evaluate the gap heat transfer coefficient in the present fuel behavior codes. The method to evaluate the gap conductance to the conditions beyond the applicability limit of the Ross and Stoute equation has also been proposed. (author)

  12. 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

  13. 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)

  14. Processing of the GALILEO fuel rod code model uncertainties within the AREVA LWR realistic thermal-mechanical analysis methodology

    International Nuclear Information System (INIS)

    Mailhe, P.; Barbier, B.; Garnier, C.; Landskron, H.; Sedlacek, R.; Arimescu, I.; Smith, M.; Bellanger, P.

    2013-01-01

    The availability of reliable tools and associated methodology able to accurately predict the LWR fuel behavior in all conditions is of great importance for safe and economic fuel usage. For that purpose, AREVA has developed its new global fuel rod performance code GALILEO along with its associated realistic thermal-mechanical analysis methodology. This realistic methodology is based on a Monte Carlo type random sampling of all relevant input variables. After having outlined the AREVA realistic methodology, this paper will be focused on the GALILEO code benchmarking process, on its extended experimental database and on the GALILEO model uncertainties assessment. The propagation of these model uncertainties through the AREVA realistic methodology is also presented. This GALILEO model uncertainties processing is of the utmost importance for accurate fuel design margin evaluation as illustrated on some application examples. With the submittal of Topical Report GALILEO to the U.S. NRC in 2013, GALILEO and its methodology are on the way to be industrially used in a wide range of irradiation conditions. (authors)

  15. 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

  16. 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

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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)

  3. 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

  4. 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

  5. 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

  6. Enhancing the ABAQUS thermomechanics code to simulate multipellet steady and transient LWR fuel rod behavior

    International Nuclear Information System (INIS)

    Williamson, R.L.

    2011-01-01

    Highlights: → The ABAQUS thermomechanics code is enhanced to enable simulation of nuclear fuel behavior. → Comparisons are made between discrete and smeared fuel pellet analysis. → Multidimensional and multipellet analysis is important for accurate prediction of PCMI. → Fully coupled thermomechanics results in very smooth prediction of fuel-clad gap closure. → A smeared-pellet approximation results in significant underprediction of clad radial displacements and plastic strain. - Abstract: A powerful multidimensional fuels performance analysis capability, applicable to both steady and transient fuel behavior, is developed based on enhancements to the commercially available ABAQUS general-purpose thermomechanics code. Enhanced capabilities are described, including: UO 2 temperature and burnup dependent thermal properties, solid and gaseous fission product swelling, fuel densification, fission gas release, cladding thermal and irradiation creep, cladding irradiation growth, gap heat transfer, and gap/plenum gas behavior during irradiation. This new capability is demonstrated using a 2D axisymmetric analysis of the upper section of a simplified multipellet fuel rod, during both steady and transient operation. Comparisons are made between discrete and smeared-pellet simulations. Computational results demonstrate the importance of a multidimensional, multipellet, fully-coupled thermomechanical approach. Interestingly, many of the inherent deficiencies in existing fuel performance codes (e.g., 1D thermomechanics, loose thermomechanical coupling, separate steady and transient analysis, cumbersome pre- and post-processing) are, in fact, ABAQUS strengths.

  7. 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

  8. 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.

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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)

  14. 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)

  15. 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.)

  16. 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)

  17. 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)

  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. 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.

  20. 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

  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. 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

  3. 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

  4. 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

  5. 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

  6. 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%.

  7. 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

  8. 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

  9. 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

  10. 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)

  11. 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

  12. 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

  13. 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

  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. 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.

  16. 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.

  17. 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

  18. 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

  19. 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.)

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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

  15. Assessment of the prediction capability of the TRANSURANUS fuel performance code on the basis of power ramp tested LWR fuel rods

    International Nuclear Information System (INIS)

    Pastore, G.; Botazzoli, P.; Di Marcello, V.; Luzzi, L.

    2009-01-01

    The present work is aimed at assessing the prediction capability of the TRANSURANUS code for the performance analysis of LWR fuel rods under power ramp conditions. The analysis refers to all the power ramp tested fuel rods belonging to the Studsvik PWR Super-Ramp and BWR Inter-Ramp Irradiation Projects, and is focused on some integral quantities (i.e., burn-up, fission gas release, cladding creep-down and failure due to pellet cladding interaction) through a systematic comparison between the code predictions and the experimental data. To this end, a suitable setup of the code is established on the basis of previous works. Besides, with reference to literature indications, a sensitivity study is carried out, which considers the 'ITU model' for fission gas burst release and modifications in the treatment of the fuel solid swelling and the cladding stress corrosion cracking. The performed analyses allow to individuate some issues, which could be useful for the future development of the code. Keywords: Light Water Reactors, Fuel Rod Performance, Power Ramps, Fission Gas Burst Release, Fuel Swelling, Pellet Cladding Interaction, Stress Corrosion Cracking

  16. 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

  17. Destructive examination of 3-cycle LWR fuel rods from Turkey Point Unit 3 for the Climax-Spent Fuel Test

    International Nuclear Information System (INIS)

    Atkin, S.D.

    1981-06-01

    The destructive examination results of five light water reactor rods from the Turkey Point Unit 3 reactor are presented. The examinations included fission gas collection and analyses, burnup and hydrogen analyses, and a metallographic evaluation of the fuel, cladding, oxide, and hydrides. The rods exhibited a low fission gas release with all other results appearing representative for pressurized water reactor fuel rods with similar burnups (28 GWd/MTU) and operating histories

  18. 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.

  19. 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

  20. 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

  1. 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

  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. 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

  4. Parameter study on the influence of prepressurization on LWR fuel rod behaviour during normal operation and hypothetical LOCA

    International Nuclear Information System (INIS)

    Fuchs, H.P.; Brzoska, B.; Depisch, F.; Sauermann, W.

    1978-01-01

    To analyse the influence of prepressurization on fuel rod behaviour, a parametric study has been performed considering the effects of the as-fabricated fuel rod internal prepressure on the normal operation and postulated LOCA red behaviour of a 1300 MWe1 KWU standard nuclear power plant pressurized water reactor. A reduction of prepressurization in the analysed range results in a negligible worsened normal operation behaviour whereas the LOCA behaviour is improved significantly. (author)

  5. PLUTON: A Three-Group Model for the Radial Distribution of Plutonium, Burnup, and Power Profiles in Highly Irradiated LWR Fuel Rods

    International Nuclear Information System (INIS)

    Lemehov, Sergei; Nakamura, Jinichi; Suzuki, Motoe

    2001-01-01

    A three-group model (PLUTON) is described, which predicts the power density distribution, plutonium buildup, and burnup profiles across the fuel pellet radius as a function of in-pile time and parameters characterizing the type of reactor system with respect to fuel temperature and changes of density during the irradiation period. The PLUTON model is a part of two fuel performance codes (ASFAD and FEMAXI-V), which provide all necessary input for this model, mainly local temperatures and fuel matrix density across the radius. Comparisons between measurements and predictions of the PLUTON model are made on fuels with enrichments in the range 2.9 to 8.25% and with burnup between 21 000 and 64 000 MWd/t. It is shown that the PLUTON predictions are in good agreement with measurements as well as with predictions of the well-known TUBRNP model. The proposed model is flexibly applicable for all types of light water reactor (LWR) fuels, including mixed oxide, and for fuel tested in the Organization for Economic Corporation and Development's Halden heavy water reactor. The PLUTON three-group model is based on analytical (theoretical) consideration of neutron absorption in a resonant region of the fuel in its apparent form. It makes the model more flexible in comparison with the semi-empirical TUBRNP one-group model and allows the physically based model analysis of commercial LWR-type fuels at high burnup as well as analysis of experimental fuel rods tested in the Halden heavy water reactor, which is one of the main test reactors in the world. The differences in fuel behavior in the Halden reactor in terms of burnup distribution and plutonium buildup can be more clearly understood with the PLUTON model

  6. 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

  7. Fluid-mixing studies in a hexagonal 217-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 217 pin LMFBR fuel bundle with a pitch to diameter ratio of 1.25 and a lead length of 12 inches. It was found that the turbulent flow data could best be characterized by the energy parameter C/sub 1L/=.106, which is 9% higher than the value from the correlation of Chiu et al. Chiu's correlation was developed on a data base of 61 and 91 pins. The spread of existing data about the correlation is +- 25%, but the error band on our data is expected to be less (approx. +- 10% since injection depth effects were not previously considered). This result is consistent with the concept of increased swirl flow in larger bundles

  8. 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)

  9. 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.

  10. 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

  11. 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.

  12. 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)

  13. 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.)

  14. HFR irradiation testing of light water reactor (LWR) fuel

    International Nuclear Information System (INIS)

    Markgraf, J.F.W.

    1985-01-01

    For the materials testing reactor HFR some characteristic information with emphasis on LWR fuel rod testing capabilities and hot cell investigation is presented. Additionally a summary of LWR fuel irradiation programmes performed and forthcoming programmes are described. Project management information and a list of publications pertaining to LWR fuel rod test programmes is given

  15. 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)

  16. ANTEO: An optimised PC computer code for the steady state thermal hydraulic analysis of rod bundles

    International Nuclear Information System (INIS)

    Cevolani, S.

    1996-07-01

    The paper deals with the description of a Personal Computer oriented subchannel code, devoted to the steady state thermal hydraulic analysis of nuclear reactor fuel bundles. The development of a such code was made possible by two facts: first, the increase the computing power of the desk machines; secondly, the fact several years of experience into operate subchannels codes have shown how to simplify many of the physical models without a sensible loss of accuracy. For sake of validation, the developed code was compared with a traditional subchannel code, the COBRA one. The results of the comparison show a very good agreement between the two codes

  17. 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

  18. 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)

  19. Operation method of the X-ray equipment for the investigation of the ballooning of LWR-fuel rod simulators

    International Nuclear Information System (INIS)

    Mueller, S.; Thun, G.

    1977-06-01

    An X-Ray-equipment is described which has been selected and assembled for the recording of fuel rod simulator-deformations during a loss of coolant accident using a movie technique. With this method it is possible to observe and record the ballooning of the simulator under conditions similar to those in a reactor. Some typical pictures are shown which show that the quality is high enough to allow a quantitative evaluation of the ballooning as a function of time. (orig.) [de

  20. SSYST, a code-system for analysing transient LWR fuel rod behaviour under off-normal conditions

    International Nuclear Information System (INIS)

    Borgwaldt, H.; Gulden, W.

    1983-01-01

    SSYST is a code-system for analysing transient fuel rod behaviour under off-normal conditions, developed conjointly by the Institut fuer Kernenergetik und Energiesysteme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract of Projek Nukleare Sicherheit (PNS) at KfK. The main differences between SSYST and similar codes are (1) an open-ended modular code organisation, and (2) a preference for simple models, wherever possible. While the first feature makes SSYST a very flexible tool, easily adapted to changing requirements, the second feature leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 min cpu-time (IBM-3033), so that extensive parametric studies become possible. This paper gives an outline of the overall code organisation and a general overview of the physical models implemented. Besides explaining the routine application of SSYST in the analysis of loss-of-coolant accidents, examples are given of special applications which have led to a satisfactory understanding of the decisive influence of deviations from rotational symmetry on the fuel rod perimeter. (author)

  1. SSYST: A code-system for analyzing transient LWR fuel rod behaviour under off-normal conditions

    International Nuclear Information System (INIS)

    Borgwaldt, H.; Gulden, W.

    1983-01-01

    SSYST is a code-system for analyzing transient fuel rod behaviour under off-normal conditions, developed conjointly by the Institut fur Kernenergetik und Energiesysteme (IKE), Stuttgart, and Kernforschungszentrum Karlsruhe (KfK) under contract of Projekt Nukleare Sicherheit (PNS) at KfK. The main differences between SSYST and similar codes are an open-ended modular code organization, and a preference for simple models, wherever possible. While the first feature makes SSYST a very flexible tool, easily adapted to changing requirements, the second feature leads to short execution times. The analysis of transient rod behaviour under LOCA boundary conditions takes 2 min cpu-time (IBM-3033), so that extensive parametric studies become possible. This paper gives an outline of the overall code organisation and a general overview of the physical models implemented. Besides explaining the routine application of SSYST in the analysis of loss-of-coolant accidents, examples are given of special applications which have led to a satisfactory understanding of the decisive influence of deviations from rotational symmetry on the fuel rod perimeter

  2. 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

  3. Stress-Strain state of structural elements of LWR fuel rods modeling in the MSC.MARC and ANSYS software

    International Nuclear Information System (INIS)

    Kuznetsov, A.; Kuznetsov, V.; Krupkin, A.; Kashirin, B.; Medvedev, A.; Novikov, V.

    2009-01-01

    The results of stress-strain state in the fuel rod spring fixing lock coils modeling are presented in this paper. The solution of this problem was realized in finite-element software MSC.MARC and ANSIS. The solution was obtained in the three-dimensional setting, taking into account multicontact interaction and all physical and geometric nonlinearities. The finite-element models were verified on analytical parities and experimental data. Results of verification have proved a correctness of the accepted finite-element models

  4. 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

  5. Fracture of Zircaloy cladding by interactions with uranium dioxide pellets in LWR fuel rods. Technical report 10

    International Nuclear Information System (INIS)

    Smith, E.; Ranjan, G.V.; Cipolla, R.C.

    1976-11-01

    Power reactor fuel rod failures can be caused by uranium dioxide fuel pellet-Zircaloy cladding interactions. The report summarizes the current position attained in a detailed theoretical study of Zircaloy cladding fracture caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn-up. It is shown that stress corrosion crack growth in irradiated Zircaloy must be able to proceed at very low stress intensifications if uniform friction effects are operative at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (i.e., ''uniform'' conditions). Otherwise, the observed fuel rod failures must be due to departures from ''uniform'' conditions, and a very high interfacial friction coefficient and particularly fuel-cladding bonding, are means of providing sufficient stess intensification at a cladding crack tip to explain the occurrence of cladding fractures. The results of the investigation focus attention on the necessity for reliable experimental data on the stress corrosion crack growth behavior of irradiated Zircaloy, and for further investigations on the correlation between local fuel-cladding bonding and stress corrosion cracking

  6. 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)

  7. The development of code for the analysis of the flow blockage of rod bundles of LMR

    International Nuclear Information System (INIS)

    Ha, Q. S.; Jeong, H. Y.; Jang, W. P.; Lee, Y. B.

    2003-01-01

    A partial flow blockage within a fuel assembly in liquid metal reactor may result in localized boiling or a failure of the fuel cladding. Thus, the precise analysis for the phenomenon is required for a safe design of LMR. To take account of the effects of the surfaces of rod and wire spacer on the fluid, the distributed resistance model was implemented into the MATRA-LMR code, which is important to the analysis for flow blockage. Also central differencing scheme for the velocities is used in the flow with the lRel less than 2 and for the enthalpies with the lPel less than 2. Diffusion terms are added to the equations of momentum and energy. The validation calculation was carried out against to the experiment of FFM series tests and the results using MATRA-LMR with the distributed resistance model and above hybrid scheme well agree with the experimental data

  8. 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)

  9. 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

  10. 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

  11. 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

  12. Interfacial area transport in two-phase flows in a scaled 8X8 rod bundle geometry at elevated pressures

    International Nuclear Information System (INIS)

    Yang, X; Schlegel, J.P.; Paranjape, S.; Liu, Y.; Chen, S.W.; Hibiki, T.; Ishii, M.

    2011-01-01

    To improve the prediction accuracy and robustness of the next-generation thermal-hydraulics system analysis code, analytical and experimental research has been undertaken to develop the Interfacial Area Transport Equation (IATE) in a scaled 8x8 rod bundle geometry at elevated pressure conditions. The experiments performed include local measurements of void fraction, interfacial area concentration, and gas velocity at several axial locations using the innovative four-sensor conductivity probe. The test conditions cover a wide range of flow regimes from bubbly, cap-bubbly, cap-turbulent to churn-turbulent at 100 kPa and 300 kPa pressure conditions and the obtained data indicates some spacer effects on the flow parameters. The bubble groups are classified into two groups (Group-1: spherical and distorted bubbles, Group-2: cap and churn turbulent bubbles) based on the bubble transport characteristics. The area-averaged interfacial area transport data have been compared to the prediction by the one-dimensional two-group IATE with mechanistically modeled IAC source and sink terms. The one-group IATE is able to predict the bubbly-flow interfacial area within ±15% error under two pressure conditions. The two-group IATE performance is also very promising in the cap-bubbly flow and churn-turbulent flow regimes, with average error of about ±20%. (author)

  13. Modelling of a rod bundle under viscous and uncompressible flow by porous media. Applied to nuclear reactor core

    International Nuclear Information System (INIS)

    Ricciardi, Guillaume; Collard, Bruno; Bellizzi, Sergio; Cochelin, Bruno

    2007-01-01

    This study is about the safety of nuclear reactor core submitted to seismic loading. In order to reduce the incertitude margin of the present day codes we propose to develop a numerical code including the non linear behavior of the fluid/structure coupling. The challenge of this work is to find out a tractable model taking the structure complexity into account. In this paper we model the nuclear reactor core mechanical behavior including the dynamics of both fuel assemblies of fluid. Each rod bundle is considered as a deformable porous media, so the velocity field of the fluid and the displacement field of the structure are defined in the whole domain space. Fluid part and structure part are in a first time considered separately, and in second time, the two parts are coupled. The motion equations of the structure are obtained by a Lagrangian formulation, and to allow the fluid structure coupling, the motion equations of the fluid are obtained by an Arbitrary Lagrangian Eulerian formulation. The finite elements method is applied to spatially discretize the equations. Simulations have been performed to analyze the influence of the fluid and structure characteristics, phenomena observed by the experience have been reproduced qualitatively. (author)

  14. 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.

  15. Critical experiment program of heterogeneous core composed for LWR fuel rods and low enriched uranyl nitrate solution

    International Nuclear Information System (INIS)

    Miyoshi, Yoshinori; Yamamoto, Toshihiro; Watanabe, Shouichi; Nakamura, Takemi

    2003-01-01

    In order to stimulate the criticality characteristics of a dissolver in a reprocessing plant, a critical experiment program of heterogeneous cores is under going at a Static Critical Experimental Facility, STACY in Japan Atomic Energy Research Institute, JAERI. The experimental system is composed of 5w/o enriched PWR-type fuel rod array immersed in 6w/o enriched uranyl nitrate solution. First series of experiments are basic benchmark experiments on fundamental critical data in order to validate criticality calculation codes for 'general-form system' classified in the Japanese Criticality Safety Handbook, JCSHB. Second series of experiments are concerning the neutron absorber effects of fission products related to the burn-up credit Level-2. For demonstrating the reactivity effects of fission products, reactivity effects of natural elements such as Sm, Nd, Eu and 103 Rh, 133 Cs, solved in the nitrate solution are to be measured. The objective of third series of experiments is to validate the effect of gadolinium as a soluble neutron poison. Properties of temperature coefficients and kinetic parameters are also studied, since these parameters are important to evaluate the transient behavior of the criticality accident. (author)

  16. 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

  17. 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.

  18. PLUTON: Three-group neutronic code for burnup analysis of isotope generation and depletion in highly irradiated LWR fuel rods

    Energy Technology Data Exchange (ETDEWEB)

    Lemehov, Sergei E; Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-08-01

    PLUTON is a three-group neutronic code analyzing, as functions of time and burnup, the change of radial profiles, together with average values, of power density, burnup, concentration of trans-uranium elements, plutonium buildup, depletion of fissile elements, and fission product generation in water reactor fuel rod with standard UO{sub 2}, UO{sub 2}-Gd{sub 2}O{sub 3}, inhomogeneous MOX, and UO{sub 2}-ThO{sub 2}. The PLUTON code, which has been designed to be run on Windows PC, has adopted a theoretical shape function of neutron attenuation in pellet, which enables users to perform a very fast and accurate calculation easily. The present code includes the irradiation conditions of the Halden Reactor which gives verification data for the code. The total list of trans-uranium elements included in the calculations consists of {sub 92}U{sup 233-239}, {sub 93}Np{sup 237-239}, {sub 94}Pu{sup 238-243}, {sub 95}Am{sup 241-244} (including isomers), and {sub 96}Cm{sup 242-245}. Poisoning fission products are represented by {sub 54}Xe{sup 131,133,135}, {sub 48}Cd{sup 113}, {sub 62}Sm{sup 149,151,152}, {sub 64}Gd{sup 154-160}, {sub 63}Eu{sup 153,155}, {sub 36}Kr{sup 83,85}, {sub 42}Mo{sup 95}, {sub 43}Tc{sup 99}, {sub 45}Rh{sup 103}, {sub 47}Ag{sup 109}, {sub 53}I{sup 127,129,131}, {sub 55}Cs{sup 133}, {sub 57}La{sup 139}, {sub 59}Pr{sup 141}, {sub 60}Nd{sup 143-150}, {sub 61}Pm{sup 147}. Fission gases and volatiles included in the code are {sub 36}Kr{sup 83-86}, {sub 54}Xe{sup 129-136}, {sub 52}Te{sup 125-130}, {sub 53}I{sup 127-131}, {sub 55}Cs{sup 133-137}, and {sub 56}Ba{sup 135-140}. Verification has been performed up to 83 GWd/tU, and a satisfactory agreement has been obtained. (author)

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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.)

  4. 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)

  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. 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.

  7. 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

  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. 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.

  10. 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.

  11. 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

  12. 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)

  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. 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

  15. Flow distribution and pressure loss in subchannels of a wire-wrapped 37-pin rod bundle for sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Seok Kyu; Euh, Dong Jin; Choi, Hae Seob; Kim, Hyung Mo; Choi, Sun Rock; Lee, Hyeong Yeon [Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and 60 degrees C (equivalent to Re ∼ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

  16. Experimental investigation of flooding in air-water counter-current flow with a vertical adiabatic multi-rod bundle

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Hho Jung; Cha, Jong Hee; Cho, Sung Jae; Chun, Moon Hyun

    1991-01-01

    The process of flooding phenomenon in a vertical adiabatic 3 x 3 tube bundle flow channel has been studied experimentally. A series of tests was performed, using three types of tube bundle differing only in the number of spacer grids attached, to investigate the effects of spacer grids and multi-flow channel interactions on the air-water counter-current flow limitations. Experimentally determined flooding points at various water film Reynolds numbers for three different test sections are presented in graphical form and compared with entrainment criterion for co-current flow and instability criteria. In addition, empirical flooding correlations of the Kutateladze type are obtained for each type of test section using liquid penetration data

  17. 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)

  18. Acoustic sensors for fission gas characterization: R and D skills devoted to innovative instrumentation in MTR, non-destructive devices in hot lab facilities and specific transducers for measurements of LWR rods in nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Ferrandis, J.Y.; Leveque, G.; Rosenkrantz, E.; Augereau, F.; Combette, P. [University Montpellier, IES, UMR 5214, F-34000, Montpellier (France); CNRS, IES, UMR 5214, F-34000, Montpellier (France)

    2015-07-01

    irradiation. The instrumented fuel rod incorporating the ultrasonic gas composition sensor was finally irradiated during 2 weeks in nominal conditions. Neutronics calculation will be performed in order to calculate the thermal and fast neutron fluence and the gamma dose absorbed by acoustic sensor. A first evaluation gives a thermal fluence about 4,5.10{sup 19} n/cm{sup 2}, a fast neutrons fluence about 4,5.1018 n/cm{sup 2} and a total gamma dose up to 0,25 MGy The maximal temperature during the irradiation test was about 150 C. Although the ultrasonic sensor appears to be damaged, the optimization of the electrical attack parameters and the development of a new signal processing maintain the measurement feasibility up the end of the irradiation campaign. It was the first time that the composition of fission gas has been monitored all along an irradiation experiment in a MTR, giving access to the gas release kinetics. New researches involve thick film transducers produced by screen-printing process in order to propose piezoelectric structures for harsh temperature and irradiation measurements. The second project consists in the development of a non-destructive device that can be directly applied on a LWR fuel rod. The problem to be solved relates to the measurement of the fission gas pressure and composition in a fuel rod using a non-destructive method. Fuel rod internal pressure is one of the safety criteria applied in nuclear power analyses. This criterion must be verified in order to avoid any fuel-cladding gap reopening risk and therefore any local clad ballooning. Apart from the safety implications, this parameter is also a fuel behaviour indicator and reflects the overall fuel performance in operation, but also during shipping and long-term storage. Rod internal pressure is one criterion amongst others, like cladding corrosion, against which the acceptable fuel burn-up limit is set. A sensor has been achieved in 2007. A full-scale hot cell test of the internal gas

  19. Recent results from CEC cost sharing research programme on LWR fuel behaviour under accident conditions

    International Nuclear Information System (INIS)

    Fairbairn, S.A.

    1983-01-01

    The present structure and intentions of the CEC sponsored cost sharing programme for LWR safety research are outlined. Detailed results are reported for two projects from this programme. The first project concerns experimental data on the thermohydraulic effects of flow diversion around ballooned fuel rods. Data are presented on single and two phase heat transfer in an electrically heated rod bundle. Detailed photographic data on droplet behaviour are also given. The second project is an investigation of the effects of zircaloy oxidation on rewetting during reflood. It is shown that as oxide thickness increases from 1μm to 76μm that rewet rates can increase by up to 40%. A systematic effect of oxidation on rewet temperatures is also noted. (author)

  20. The initial study on supercritical water flow and heat transfer in square rod bundle channel with mixing vane

    International Nuclear Information System (INIS)

    Zuo Guoping; Cao Can; Yu Tao

    2010-01-01

    Three-dimensional rectangular channel with the mixing wine in supercritical water reactor was studied in the paper using the FLUENT software. The mixing wing elevation influence on temperature distribution and flow field were studied in the model. The results showed the mixing wing caused fluid circumferential flow, making flow hot and cold fluids mixed and fluid temperature uniform distribution, effectively improved the fuel rod surface temperature distribution and reduced hot temperature. Among the four cases of mixing wing elevation of 15, 30, 45 and 50 angle, 30 angle is the best case in improving temperature distribution. (authors)

  1. 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

  2. Continual approach to the dynamics problems of tanks containing rod bundles or particle groups and fluid at vibrational actions

    International Nuclear Information System (INIS)

    Fedotovskii, V.S.

    1988-02-01

    The vibration of tanks with liquid and non deformed cylindrical or spherical inclusions are considered. It is shown that for calculating dynamic characteristics of such systems it is advisable to use continual approach i.e. consider-heterogeneous media formed by liquid and weighted inclusions in it as homogeneous media with effective or vibroreological properties. On the base of the problem on vibrations of the tank, containing liquid and localized inclusions, rod assemblies vibrations are considered and relationships for the added mass and resistance coefficient determining dynamic characteristics of such systems are obtained. Considered are also liquid tank vibrations containing spherical inclusions. The results obtained are used for calculating dynamic characteristics of two-phase flow pipelines at bubble and annular flow mode. The theoretical relationships are compared with available experimental data [fr

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  4. 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)

  5. Effects of reduced surface tension on two-phase diversion cross-flow between subchannels simplifying triangle tight lattice rod bundle

    International Nuclear Information System (INIS)

    Kawahara, Akimaro; Sadatomi, Michio; Higuchi, Tatsuya

    2009-01-01

    Two-phase diversion cross-flow between tight lattice subchannels has been investigated experimentally and analytically. For hydraulically non-equilibrium flows with the pressure difference between the subchannels, experiments were conducted using a vertical multiple-channel with two subchannels simplifying a triangle tight lattice rod bundle. To know the effects of the reduced surface tension on the diversion cross-flow, water and water with a surfactant were used as the test liquids. Data were obtained on the axial variations in the pressure difference between the subchannels, gas and liquid flow rates and void fraction in each subchannel for slug-churn and annular flows. In the analysis, flow redistribution processes due to the diversion cross-flow have been calculated by our subchannel analysis code based on a two-fluid model. From a comparison between the experiment and the code calculation, the code was found to be valid against the present data if the improved constitutive equations of wall and interfacial friction reported in our previous paper were incorporated to account for the reduced surface tension effects. (author)

  6. Simulation of single-phase rod bundle flow. Comparison between CFD-code ESTET, PWR core code THYC and experimental results

    International Nuclear Information System (INIS)

    Mur, J.; Larrauri, D.

    1998-07-01

    Computer simulation of flow in configurations close to pressurized water reactor (PWR) geometry is of great interest for Electricite de France (EDF). Although simulation of the flow through a whole PWR core with an all purpose CFD-code is not yet achievable, such a tool cna be quite useful to perform numerical experiments in order to try and improve the modeling introduced in computer codes devoted to reactor core thermal-hydraulic analysis. Further to simulation in small bare rod bundle configurations, the present study is focused on the simulation, with CFD-code ESTET and PWR core code THYC, of the flow in the experimental configuration VATICAN-1. ESTET simulation results are compared on the one hand to local velocity and concentration measurements, on the other hand with subchannel averaged values calculated by THYC. As far as the comparison with measurements is concerned, ESTET results are quite satisfactory relatively to available experimental data and their uncertainties. The effect of spacer grids and the prediction of the evolution of an unbalanced velocity profile seem to be correctly treated. As far as the comparison with THYC subchannel averaged values is concerned, the difficulty of a direct comparison between subchannel averaged and local values is pointed out. ESTET calculated local values are close to experimental local values. ESTET subchannel averaged values are also close to THYC calculation results. Thus, THYC results are satisfactory whereas their direct comparison to local measurements could show some disagreement. (author)

  7. Parametric Evaluation of SiC/SiC Composite Cladding with UO2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

    Science.gov (United States)

    Singh, G.; Sweet, R.; Brown, N. R.; Wirth, B. D.; Katoh, Y.; Terrani, K.

    2018-02-01

    SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. To ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel-cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

  8. 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.

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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

  14. Experimental study of fuel bundle vibrations with rods subjected to mixed axial flow and cross-flow provided by a narrow gap (baffle jetting interaction)

    International Nuclear Information System (INIS)

    Boulanger, P.; Jacques, Y.; Fardeau, P.; Barbier, D.; Rigaudeau, J.

    1997-01-01

    The Hydraulic Core Laboratory (LHC) performs experimental studies of PWR fuel assembly mechanical behaviour submitted to representative flows in PWR core. Cross-flows prove particularly troublesome by generating on rods, in special cases, vibratory levels high enough to induce early grid to rod fretting. The fluid-structure interaction under mixed axial and cross-flow is also a major topic for analysis. The authors present a test loop devoted to the mixed axial-cross-flow fluid-structure interaction on representative half-scale mockup which is able to simulate, under ambient conditions, any complex flow (direction and flow rates) representative of PWR core flows. Despite its reduced size, the mockup retains the overall structure of a PWR fuel assembly. Rods displacement/velocity and velocity flow field are measured by laser techniques

  15. 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

  16. Multi-dimensional modeling of two-phase flow in rod bundles and interpretation of velocities measured in BWRs by the cross-correlation technique

    International Nuclear Information System (INIS)

    Analytis, G.Th.; Luebbesmeyer, D.

    1984-04-01

    The authors present an as precise as possible interpretation of velocity measurements in BWRs by the cross-correlation technique, which is based on the radially non-uniform quality and velocity distribution in BWR type bundles, as well as on our knowledge about the spatial 'field of view' of the in-core neutron detectors. After formulating the three-dimensional two-fluid model volume/time averaged equations and pointing out some problems associated with averaging, they expound a little on the turbulence mixing and void drift effects, as well as on the way they are modelled in advanced subchannel analysis codes like THERMIT or COBRA-TF. Subsequently, some comparisons are made between axial velocities measured in a commercial BWR by neutron noise analysis, and the steam velocities of the four subchannels nearest to the instrument tube of one of the four bundles as predicted by COBRA-III and by THERMIT. Although as expected, for well-known reasons, COBRA-III predicts subchannel steam velocities which are close to each other, THERMIT correctly predicts in the upper half of the core three largely different steam velocities in the three different types of BW0 subchannels (corner, edge and interior). (Auth.)

  17. 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.)

  18. An analytical model for the prediction of fluid-elastic forces in a rod bundle subjected to axial flow: theory, experimental validation and application to PWR fuel assemblies

    International Nuclear Information System (INIS)

    Beaud, F.

    1997-01-01

    A model predicting the fluid-elastic forces in a bundle of circular cylinders subjected to axial flow is presented in this paper. Whereas previously published models were limited to circular flow channel, the present one allows to take a rectangular flow external boundary into account. For that purpose, an original approach is derived from the standard method of images. This model will eventually be used to predict the fluid-structure coupling between the flow of primary coolant and a fuel assemblies in PWR nuclear reactors. It is indeed of major importance since the flow is shown to induce quite high damping and could therefore mitigate the incidence of an external load like a seismic excitation on the dynamics of the assemblies. The proposed model is validated on two cases from the literature but still needs further comparisons with the experiments being currently carried out on the EDF set-up. The flow has been shown to induce an approximate 12% damping on a PWR fuel assembly, at nominal reactor conditions. The possible grid effect on the fluid-structure coupling has been neglected so far but will soon be investigated at EDF. (author)

  19. Polyelectrolyte bundles

    Energy Technology Data Exchange (ETDEWEB)

    Limbach, H J; Sayar, M; Holm, C [Max-Planck-Institut fuer Polymerforschung, Ackermannweg 10, 55128 Mainz (Germany)

    2004-06-09

    Using extensive molecular dynamics simulations we study the behaviour of polyelectrolytes with hydrophobic side chains, which are known to form cylindrical micelles in aqueous solution. We investigate the stability of such bundles with respect to hydrophobicity, the strength of the electrostatic interaction and the bundle size. We show that for the parameter range relevant for sulfonated poly(para-phenylenes) (PPP) one finds a stable finite bundle size. In a more generic model we also show the influence of the length of the precursor oligomer on the stability of the bundles. We also point out that our model has close similarities to DNA solutions with added condensing agents, hinting at the possibility that the size of DNA aggregates is, under certain circumstances, thermodynamically limited.

  20. Polyelectrolyte bundles

    International Nuclear Information System (INIS)

    Limbach, H J; Sayar, M; Holm, C

    2004-01-01

    Using extensive molecular dynamics simulations we study the behaviour of polyelectrolytes with hydrophobic side chains, which are known to form cylindrical micelles in aqueous solution. We investigate the stability of such bundles with respect to hydrophobicity, the strength of the electrostatic interaction and the bundle size. We show that for the parameter range relevant for sulfonated poly(para-phenylenes) (PPP) one finds a stable finite bundle size. In a more generic model we also show the influence of the length of the precursor oligomer on the stability of the bundles. We also point out that our model has close similarities to DNA solutions with added condensing agents, hinting at the possibility that the size of DNA aggregates is, under certain circumstances, thermodynamically limited

  1. Polyelectrolyte bundles

    Science.gov (United States)

    Limbach, H. J.; Sayar, M.; Holm, C.

    2004-06-01

    Using extensive Molecular Dynamics simulations we study the behavior of polyelectrolytes with hydrophobic side chains, which are known to form cylindrical micelles in aqueous solution. We investigate the stability of such bundles with respect to hydrophobicity, the strength of the electrostatic interaction, and the bundle size. We show that for the parameter range relevant for sulfonated poly-para-phenylenes (PPP) one finds a stable finite bundle size. In a more generic model we also show the influence of the length of the precursor oligomer on the stability of the bundles. We also point out that our model has close similarities to DNA solutions with added condensing agents, hinting to the possibility that the size of DNA aggregates is under certain circumstances thermodynamically limited.

  2. 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)

  3. 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)

  4. Qualification of ARROTTA code for LWR accident analysis

    International Nuclear Information System (INIS)

    Huang, P.-H.; Peng, K.Y.; Lin, W.-C.; Wu, J.-Y.

    2004-01-01

    This paper presents the qualification efforts performed by TPC and INER for the 3-D spatial kinetics code ARROTTA for LWR core transient analysis. TPC and INER started a joint 5 year project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multi-dimensional kinetics calculations such as rod ejection for PWR and rod drop for BWR. To qualify ARROTTA for analysis of FSAR licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicated that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include NEACRP rod ejection problem, 3-D LMW LWR rod withdrawal/insertion problem, and 3-D LRA BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared to other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multi-dimensional core transient for LWRs. (author)

  5. 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

  6. 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)

  7. Ultrasonically identified cap seal for LWR fuel bundles

    International Nuclear Information System (INIS)

    Buergers, W.; Dal Cero, J.; Crutzen, S.

    1981-01-01

    This paper aims to provide a general review of techniques available for surveillance and for sealing, marking or otherwise identifying material in such a way that its recognition and guarantee of integrity are unequivocally assured. The problem of obtaining such assurance has been the subject of work at Ispra and elsewhere. Some discussion of the problems, the work performed and possible solution are given. In addition, techniques which, although not yet in routine use, would be suitable for such application are described. Using industrial ultrasonic apparatus, the signal obtained by scanning the seals was very satisfactory, as was shown by the evaluation studies. The general method is based on reflection due to the great difference in acoustical impedance existing between the matrix and the inclusions

  8. 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)

  9. Light Water Reactor (LWR) safety

    International Nuclear Information System (INIS)

    Sehgal, Bal Raj

    2006-01-01

    In this paper, a historical review of the developments in the safely of LWR power plants is presented. The paper reviews the developments prior to the TMI-2 accident, i.e. the concept of the defense in depth, the design basis, the large LOCA technical controversies and the LWR safety research programs. The TMI-2 accident, which became a turning point in the history of the development of nuclear power is described briefly. The Chernobyl accident, which terrified the world and almost completely curtailed the development of nuclear power is also described briefly. The great international effort of research in the LWR design-base and severe accidents, which was, respectively, conducted prior to and following the TMI-2 and Chernobyl accidents is described next. We conclude that with the knowledge gained and the improvements in plant organisation/management and in the training of the staff at the presently-installed nuclear power stations, the LWR plants have achieved very high standards of safety and performance. The Generation 3 + LWR power plants, next to be installed, may claim to have reached the goal of assuring the safety of the public to a very large extent. This review is based on the historical developments in LWR safety that occurred primarily in USA. however, they are valid for the rest of the Western World. This review can not do justice to the many many fine contributions that have been made over the last fifty years to the cause of LWR safety. We apologize if we have not mentioned them. We also apologize for not providing references to many of the fine investigations, which have contributed towards LWR safety earning the conclusions that we describe just above

  10. Control rod

    International Nuclear Information System (INIS)

    Kawakami, Kazuo; Shimoshige, Takanori; Nishimura, Akira

    1979-01-01

    Purpose: A control rod has been developed, which provided a plurality of through-holes in the vicinity of the sheath fitting position, in order to flatten burn-up, of fuel rods in positions confronting a control rod. Thereby to facilitate the manufacture of the control rods and prevent fuel rod failures. Constitution: A plurality of through-holes are formed in the vicinity of the sheath fitting position of a central support rod to which a sheath for the control rod is fitted. These through-holes are arranged in the axial direction of the central support rod. Accordingly, burn-up of fuel rods confronting the control rods can be reduced by through-holes and fuel rod failures can be prevented. (Yoshino, Y.)

  11. Irradiation effects on thermal properties of LWR hydride fuel

    Energy Technology Data Exchange (ETDEWEB)

    Terrani, Kurt, E-mail: terrani@berkeley.edu [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Balooch, Mehdi [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States); Carpenter, David; Kohse, Gordon [Massachusetts Institute of Technology, 138 Albany St., Cambridge, MA 02139 (United States); Keiser, Dennis; Meyer, Mitchell [Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Olander, Donald [University of California, 4155 Etcheverry Hall, M.C. 1730, Berkeley, CA 94720-1730 (United States)

    2017-04-01

    Three hydride mini-fuel rods were fabricated and irradiated at the MIT nuclear reactor with a maximum burnup of 0.31% FIMA or ∼5 MWd/kgU equivalent oxide fuel burnup. Fuel rods consisted of uranium-zirconium hydride (U (30 wt%)ZrH{sub 1.6}) pellets clad inside a LWR Zircaloy-2 tubing. The gap between the fuel and the cladding was filled with lead-bismuth eutectic alloy to eliminate the gas gap and the large temperature drop across it. Each mini-fuel rod was instrumented with two thermocouples with tips that are axially located halfway through the fuel centerline and cladding surface. In-pile temperature measurements enabled calculation of thermal conductivity in this fuel as a function of temperature and burnup. In-pile thermal conductivity at the beginning of test agreed well with out-of-pile measurements on unirradiated fuel and decreased rapidly with burnup.

  12. Development of LWR fuel performance code FEMAXI-6

    International Nuclear Information System (INIS)

    Suzuki, Motoe

    2006-01-01

    LWR fuel performance code: FEMAXI-6 (Finite Element Method in AXIs-symmetric system) is a representative fuel analysis code in Japan. Development history, background, design idea, features of model, and future are stated. Characteristic performance of LWR fuel and analysis code, what is model, development history of FEMAXI, use of FEMAXI code, fuel model, and a special feature of FEMAXI model is described. As examples of analysis, PCMI (Pellet-Clad Mechanical Interaction), fission gas release, gap bonding, and fission gas bubble swelling are reported. Thermal analysis and dynamic analysis system of FEMAXI-6, function block at one time step of FEMAXI-6, analytical example of PCMI in the output increase test by FEMAXI-III, analysis of fission gas release in Halden reactor by FEMAXI-V, comparison of the center temperature of fuel in Halden reactor, and analysis of change of diameter of fuel rod in high burn up BWR fuel are shown. (S.Y.)

  13. 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)

  14. 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

  15. 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

  16. Overview of P.I.E. techniques for L.W.R. fuels at Saclay hot cells with special emphasis on new apparatus and on mechanical testing

    International Nuclear Information System (INIS)

    Blanc, J.Y.; Hardy, J.L.; Trotabas, M.

    1990-01-01

    This paper describes the state-of-the-art in the Saclay hot cells for examining L.W.R. fuels. First, we present the classical path followed by a fuel rod in the laboratory, to begin with non-destructive testing. This is completed by destructive examinations, such as free volume determination and fission gases analyses, density measurement and metallographies including X-rays diffraction and microprobe (EPMA/WDX). These two last techniques enable the identification of elements and chemical nature of compounds which are present. We also perform mechanical tests on metallic components, on clads and guide-tubes (tensile tests, creep, burst or fatigue tests by internal pressure). Another apparatus is devoted to the study of irradiated clad behaviour during LOCA-type transients. In the second chapter, a particular emphasis is given to the developments in progress, or planned in the near future. This includes: (a) The implementation of a new non-destructive testing bench to inspect more fuel rods simultaneously. (b) A new image analyzer to be applied e.g. to hydrides analysis in the clad, or to the inspection of safety test fuel bundles. (c) As for mechanical testing, we describe here the tensile tests on clads or on guide-tubes, performed on longitudinal samples or ring samples

  17. Effects of cooling time on a closed LWR fuel cycle

    International Nuclear Information System (INIS)

    Arnold, R. P.; Forsberg, C. W.; Shwageraus, E.

    2012-01-01

    In this study, the effects of cooling time prior to reprocessing spent LWR fuel has on the reactor physics characteristics of a PWR fully loaded with homogeneously mixed U-Pu or U-TRU oxide (MOX) fuel is examined. A reactor physics analysis was completed using the CASM04e code. A void reactivity feedback coefficient analysis was also completed for an infinite lattice of fresh fuel assemblies. Some useful conclusions can be made regarding the effect that cooling time prior to reprocessing spent LWR fuel has on a closed homogeneous MOX fuel cycle. The computational analysis shows that it is more neutronically efficient to reprocess cooled spent fuel into homogeneous MOX fuel rods earlier rather than later as the fissile fuel content decreases with time. Also, the number of spent fuel rods needed to fabricate one MOX fuel rod increases as cooling time increases. In the case of TRU MOX fuel, with time, there is an economic tradeoff between fuel handling difficulty and higher throughput of fuel to be reprocessed. The void coefficient analysis shows that the void coefficient becomes progressively more restrictive on fuel Pu content with increasing spent fuel cooling time before reprocessing. (authors)

  18. Control rods

    International Nuclear Information System (INIS)

    Maruyama, Hiromi.

    1984-01-01

    Purpose: To realize effective utilization, cost reduction and weight reduction in neutron absorbing materials. Constitution: Residual amount of neutron absorbing material is averaged between the top end region and other regions of a control rod upon reaching to the control rod working life, by using a single kind of neutron absorbing material and increasing the amount of the neutron absorber material at the top end region of the control rod as compared with that in the other regions. Further, in a case of a control rod having control rod blades such as in a cross-like control rod, the amount of the neutron absorbing material is decreased in the middle portion than in the both end portions of the control rod blade along the transversal direction of the rod, so that the residual amount of the neutron absorbing material is balanced between the central region and both end regions upon reaching the working life of the control rod. (Yoshihara, H.)

  19. Fission product release from high gap-inventory LWR fuel under LOCA conditions

    International Nuclear Information System (INIS)

    Lorenz, R.A.; Collins, J.L.; Osborne, M.F.; Malinauskas, A.P.

    1980-01-01

    Fission product release tests were performed with light water reactor (LWR) fuel rod segments containing large amounts of cesium and iodine in the pellet-to-cladding gap space in order to check the validity of the previously published Source Term Model for this type of fuel. The model describes the release of fission product cesium and iodine from LWR fuel rods for controlled loss-of-coolant accident (LOCA) transients in the temperature range 500 to 1200 0 C. The basis for the model was test data obtained with simulated fuel rods and commercial fuel irradiated to high burnup but containing relatively small amounts of cesium and iodine in the pellet-to-cladding gap space

  20. Integrity of neutron-absorbing components of LWR fuel systems

    International Nuclear Information System (INIS)

    Bailey, W.J.; Berting, F.M.

    1991-03-01

    A study of the integrity and behavior of neutron-absorbing components of light-water (LWR) fuel systems was performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE). The components studies include control blades (cruciforms) for boiling-water reactors (BWRs) and rod cluster control assemblies for pressurized-water reactors (PWRs). The results of this study can be useful for understanding the degradation of neutron-absorbing components and for waste management planning and repository design. The report includes examples of the types of degradation, damage, or failures that have been encountered. Conclusions and recommendations are listed. 84 refs

  1. LWR mox fuel experience in Belgium and France with special emphasis on results obtained in BR3

    International Nuclear Information System (INIS)

    Bairiot, H.; Haas, D.; Lippens, M.; Motte, F.; Lebastard, G.; Marin, J.F.

    1986-09-01

    The course of the paper reflects two main topics: LWR MOX fuel experience in Belgium and France, summarizing the fabrication techniques, the references, the underlying MOX fuel technology and the current R and D programs for expanding the data base; behaviour of MOX fuel rods irradiated under steady state and transient operating conditions, focusing on MOX fuel technology features acquired through the irradiations performed in the BR3 PWR, supplemented by tests in the BR2 MTR. This paper focuses on the thermomechanical behaviour of LWR MOX fuel rods, which is intimately related to the fabrication technique and vice-versa. 22 refs

  2. LWR-core behaviour project

    International Nuclear Information System (INIS)

    Paratte, J.M.

    1982-07-01

    The LWR-Core behaviour project concerns the mathematical simulation of a light water reactor in normal operation (emergency situations excluded). Computational tools are assembled, i.e. programs and libraries of data. These computational tools can likewise be used in nuclear power applications, industry and control applications. The project is divided into three parts: the development and application of calculation methods for quantisation determination of LWR physics; investigation of the behaviour of nuclear fuels under radiation with special attention to higher burnup; simulation of the operating transients of nuclear power stations. (A.N.K.)

  3. Control rod

    International Nuclear Information System (INIS)

    Igarashi, Takao; Sugawara, Satoshi; Yoshimoto, Yuichiro; Saito, Shozo; Fukumoto, Takashi.

    1987-01-01

    Purpose: To reduce the weight and thereby obtain satisfactory operationability of control rods by combining absorbing nuclear chain type neutron absorbers and conventional type neutron absorbers in the axial direction of blades. Constitution: Neutron absorber rods and long life type neutron absorber rods are disposed in a tie rod and a sheath. The neutron absorber rod comprises a poison tube made of stainless steels and packed with B 4 C powder. The long life type neutron absorber rod is prepared by packing B-10 enriched boron carbide powder into a hafnium metal rod, hafnium pipe, europium and stainless made poison tube. Since the long life type absorber rod uses HF as the absorbing nuclear chain type neutron absorber, it absorbs neutrons to form new neutron absorbers to increase the nuclear life. (Yoshino, Y.)

  4. A simplified computational scheme for thermal analysis of LWR spent fuel dry storage and transportation casks

    International Nuclear Information System (INIS)

    Kim, Chang Hyun

    1997-02-01

    A simplified computational scheme for thermal analysis of the LWR spent fuel dry storage and transportation casks has been developed using two-step thermal analysis method incorporating effective thermal conductivity model for the homogenized spent fuel assembly. Although a lot of computer codes and analytical models have been developed for application to the fields of thermal analysis of dry storage and/or transportation casks, some difficulties in its analysis arise from the complexity of the geometry including the rod bundles of spent fuel and the heat transfer phenomena in the cavity of cask. Particularly, if the disk-type structures such as fuel baskets and aluminium heat transfer fins are included, the thermal analysis problems in the cavity are very complex. To overcome these difficulties, cylindrical coordinate system is adopted to calculate the temperature profile of a cylindrical cask body using the multiple cylinder model as the step-1 analysis of the present study. In the step-2 analysis, Cartesian coordinate system is adopted to calculate the temperature distributions of the disk-type structures such as fuel basket and aluminium heat transfer fin using three- dimensional conduction analysis model. The effective thermal conductivity for homogenized spent fuel assembly based on Manteufel and Todreas model is incorporated in step-2 analysis to predict the maximum fuel temperature. The presented two-step computational scheme has been performed using an existing HEATING 7.2 code and the effective thermal conductivity for the homogenized spent fuel assembly has been calculated by additional numerical analyses. Sample analyses of five cases are performed for NAC-STC including normal transportation condition to examine the applicability of the presented simplified computational scheme for thermal analysis of the large LWR spent fuel dry storage and transportation casks and heat transfer characteristics in the cavity of the cask with the disk-type structures

  5. 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

  6. 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

  7. 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

  8. A study on the behavior of defected LWR spent fuel

    International Nuclear Information System (INIS)

    You, Gil Sung; Kim, Eun Ka; Kim, Keon Sik; Suh, Hang Suck; Kim, Seung Jung; Ro, Seung Gy; Park, Chong Mook; Ji, Pyung Gook

    1992-03-01

    To investigate the storage behavior of the defective LWR spent fuel rods, the characteristic changes of fuel and cladding are to be measured and analyzed. In addition, the oxidation study in air on non-irradiated and irradiated U0 2 was performed. No changes were observed in the tested fuel rods after 30 month storage. The Cs-134, 137 released rapidly during the initial 3 months of storage, but remained in constant value after 3 month storage and the release was almost ceased after 30 month storage. The weight gain of non-irradiated U0 2 samples showed a trend of S type curves and the activation energies were 11OKJ/mol above 350 deg C. and 143KJ/mol below 350 deg C. But irradiated U0 2 showed a rapid increase at initial stage of oxidation and a decrease at later stage when compared with the results of non-irradiated U0 2 . (Author)

  9. Validating the BISON fuel performance code to integral LWR experiments

    Energy Technology Data Exchange (ETDEWEB)

    Williamson, R.L., E-mail: Richard.Williamson@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gamble, K.A., E-mail: Kyle.Gamble@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Perez, D.M., E-mail: Danielle.Perez@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Novascone, S.R., E-mail: Stephen.Novascone@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Pastore, G., E-mail: Giovanni.Pastore@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Gardner, R.J., E-mail: Russell.Gardner@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Hales, J.D., E-mail: Jason.Hales@inl.gov [Fuel Modeling and Simulation, Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-3840 (United States); Liu, W., E-mail: Wenfeng.Liu@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States); Mai, A., E-mail: Anh.Mai@anatech.com [ANATECH Corporation, 5435 Oberlin Dr., San Diego, CA 92121 (United States)

    2016-05-15

    Highlights: • The BISON multidimensional fuel performance code is being validated to integral LWR experiments. • Code and solution verification are necessary prerequisites to validation. • Fuel centerline temperature comparisons through all phases of fuel life are very reasonable. • Accuracy in predicting fission gas release is consistent with state-of-the-art modeling and the involved uncertainties. • Rod diameter comparisons are not satisfactory and further investigation is underway. - Abstract: BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison

  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. 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

  12. Technical development on burn-up credit for spent LWR fuels

    International Nuclear Information System (INIS)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori

    2000-10-01

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

  13. Technical Development on Burn-up Credit for Spent LWR Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gauld, I.C.

    2001-12-26

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled ''Technical Development on Criticality Safety Management for Spent LWR Fuels.'' Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burnup and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report.

  14. Technical development on burn-up credit for spent LWR fuels

    Energy Technology Data Exchange (ETDEWEB)

    Nakahara, Yoshinori; Suyama, Kenya; Suzaki, Takenori [eds.] [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2000-10-01

    Technical development on burn-up credit for spent LWR fuels had been performed at JAERI since 1990 under the contract with Science and Technology Agency of Japan entitled 'Technical Development on Criticality Safety Management for Spent LWR Fuels'. Main purposes of this work are to obtain the experimental data on criticality properties and isotopic compositions of spent LWR fuels and to verify burn-up and criticality calculation codes. In this work three major experiments of exponential experiments for spent fuel assemblies to obtain criticality data, non-destructive gamma-ray measurement of spent fuel rods for evaluating axial burn-up profiles, and destructive analyses of spent fuel samples for determining precise burn-up and isotopic compositions were carried out. The measured data obtained were used for validating calculation codes as well as an examination of criticality safety analyses. Details of the work are described in this report. (author)

  15. Nonabelian bundle 2-gerbes

    OpenAIRE

    Jurco, Branislav

    2009-01-01

    We define 2-crossed module bundle 2-gerbes related to general Lie 2-crossed modules and discuss their properties. A 2-crossed module bundle 2-gerbe over a manifold is defined in terms of a so called 2-crossed module bundle gerbe, which is a crossed module bundle gerbe equipped with an extra sructure. It is shown that string structures can be described and classified using 2-crossed module bundle 2-gerbes.

  16. Replacement rod

    International Nuclear Information System (INIS)

    Hatfield, S.C.

    1989-01-01

    This patent describes in an elongated replacement rod for use with fuel assemblies of the type having two end fittings connected by guide tubes with a plurality of rod and guide tube cell defining spacer grids containing rod support features and mixing vanes. The grids secured to the guide tubes in register between the end fittings at spaced intervals. The fuel rod comprising: an asymmetrically beveled tip; a shank portion having a straight centerline; and a permanently diverging portion between the tip and the shank portion

  17. 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)

  18. 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.

  19. ERDA LWR plant technology program: role of government/industry in improving LWR performance

    International Nuclear Information System (INIS)

    1975-01-01

    Information is presented under the following chapter headings: executive summary; LWR plant outages; LWR plant construction delays and cancellations; programs addressing plant outages, construction delays, and cancellations; need for additional programs to remedy continuing problems; criteria for government role in LWR commercialization; and the proposed government program

  20. Geometric radiation exchange factors for axial radiative transfer in an LWR core filled with absorbing-emitting gases

    International Nuclear Information System (INIS)

    Chan, S.H.; Cho, D.H.

    1984-01-01

    A reactor core filled with an emitting-absorbing mixture (like steam, hydrogen gas and fission gases) is considered. Analysis is provided to evaluate axial radiative heat exchange of a rod bundle with a nonuniform axial temperature distribution. The necessary radiation exchange shape factors (geometric mean absorptance, emittance and transmittance) between segments of the complex rod bundle arrangement are presented. They are applicable to arbitrary sizes of segments, well suited for numerical computations

  1. Water rod

    International Nuclear Information System (INIS)

    Kashiwai, Shin-ichi; Yokomizo, Osamu; Orii, Akihito.

    1992-01-01

    In a reactor core of a BWR type reactor, the area of a flow channel in a lower portion of a downcoming pipe for downwardly releasing steams present at the top portion in a water rod is increased. Further, a third coolant flow channel (an inner water rod) is disposed in an uprising having an exit opened near the inlet of the water rod and an inlet opened at the outside near the top portion of the water and having an increase flow channel area in the upper portion. The downcoming pipe in the water rod is filled with steams, and the void ratio is increased by so much as the flow channel area of the downcoming pipe is increased. Since the pressure difference between the inlet and the exit of the inner water rod is greater than the pressure difference between the inlet and the exit of the water rod, most of water flown into the inner water rod is discharged out of the exit in the form of water as it is. Since the area of the flow channel is increased in the portion of the inner water rod, void efficiency in the upper portion of the reactor core is decreased by so much. Since the void ratio is thus increased in the lower portion and the void efficiency is decreased in the upper portion of the reactor core, axial void distribution can be flattened. (N.H.)

  2. Results of international standard problem No. 36 severe fuel damage experiment of a VVER fuel bundle

    Energy Technology Data Exchange (ETDEWEB)

    Firnhaber, M. [Gesellschaft fuer Anlagen-und Reaktorsicherheit, Koeln (Germany); Yegorova, L. [Nuclear Safety Institute of Russian Research Center, Moscow (Russian Federation); Brockmeier, U. [Ruhr-Univ. of Bochum (Germany)] [and others

    1995-09-01

    International Standard Problems (ISP) organized by the OECD are defined as comparative exercises in which predictions with different computer codes for a given physical problem are compared with each other and with a carefully controlled experimental study. The main goal of ISP is to increase confidence in the validity and accuracy of analytical tools used in assessing the safety of nuclear installations. In addition, it enables the code user to gain experience and to improve his competence. This paper presents the results and assessment of ISP No. 36, which deals with the early core degradation phase during an unmitigated severe LWR accident in a Russian type VVER. Representatives of 17 organizations participated in the ISP using the codes ATHLET-CD, ICARE2, KESS-III, MELCOR, SCDAP/RELAP5 and RAPTA. Some participants performed several calculations with different codes. As experimental basis the severe fuel damage experiment CORA-W2 was selected. The main phenomena investigated are thermal behavior of fuel rods, onset of temperature escalation, material behavior and hydrogen generation. In general, the calculations give the right tendency of the experimental results for the thermal behavior, the hydrogen generation and, partly, for the material behavior. However, some calculations deviate in important quantities - e.g. some material behavior data - showing remarkable discrepancies between each other and from the experiments. The temperature history of the bundle up to the beginning of significant oxidation was calculated quite well. Deviations seem to be related to the overall heat balance. Since the material behavior of the bundle is to a great extent influenced by the cladding failure criteria a more realistic cladding failure model should be developed at least for the detailed, mechanistic codes. Regarding the material behavior and flow blockage some models for the material interaction as well as for relocation and refreezing requires further improvement.

  3. Strategic Aspects of Bundling

    International Nuclear Information System (INIS)

    Podesta, Marion

    2008-01-01

    The increase of bundle supply has become widespread in several sectors (for instance in telecommunications and energy fields). This paper review relates strategic aspects of bundling. The main purpose of this paper is to analyze profitability of bundling strategies according to the degree of competition and the characteristics of goods. Moreover, bundling can be used as price discrimination tool, screening device or entry barriers. In monopoly case bundling strategy is efficient to sort consumers in different categories in order to capture a maximum of surplus. However, when competition increases, the profitability on bundling strategies depends on correlation of consumers' reservations values. (author)

  4. 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.)

  5. The dupic fuel cycle synergism between LWR and HWR

    International Nuclear Information System (INIS)

    Lee, J.S.; Yang, M.S.; Park, H.S.; Lee, H.H.; Kim, K.P.; Sullivan, J.D.; Boczar, P.G.; Gadsby, R.D.

    1999-01-01

    The DUPIC fuel cycle can be developed as an alternative to the conventional spent fuel management options of direct disposal or plutonium recycle. Spent LWR fuel can be burned again in a HWR by direct refabrication into CANDU-compatible DUPIC fuel bundles. Such a linkage between LWR and HWR can result in a multitude of synergistic effects, ranging from savings of natural uranium to reductions in the amount of spent fuel to be buried in the earth, for a given amount of nuclear electricity generated. A special feature of the DUPIC fuel cycle is its compliance with the 'Spent Fuel Standard' criteria for diversion resistance, throughout the entire fuel cycle. The DUPIC cycle thus has a very high degree of proliferation resistance. The cost penalty due to this technical factor needs to be considered in balance with the overall benefits of the DUPIC fuel cycle. The DUPIC alternative may be able to make a significant contribution to reducing spent nuclear fuel burial in the geosphere, in a manner similar to the contribution of the nuclear energy alternative in reducing atmospheric pollution from fossil fuel combustion. (author)

  6. 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)

  7. Bundle Branch Block

    Science.gov (United States)

    ... known cause. Causes can include: Left bundle branch block Heart attacks (myocardial infarction) Thickened, stiffened or weakened ... myocarditis) High blood pressure (hypertension) Right bundle branch block A heart abnormality that's present at birth (congenital) — ...

  8. 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

  9. Recycling U and Pu in LWR

    International Nuclear Information System (INIS)

    Zheng Hualing.

    1986-01-01

    This article, from viewpoints of technical feasibility, safety evaluation and socioeconomic benefit-risk analysis, introduces and comments on history and status of recycling U and Pu in LWR, dealing with reactor, reprocessing, conversion and fuel element fabrication et al. Author has analysed LWR fuel cycle strategies in China and made a proposal

  10. LWR physics in SKODA Works

    International Nuclear Information System (INIS)

    Zbytovsky, A.; Lehmann, M.; Vyskocil, V.; Vacek, J.; Krysl, V.

    1980-01-01

    Computation of nuclear power reactors of the WWER-1000 type is described as are computer programs used by Skoda Works for the solution of neutron problems. The programs are analyzed for applicability in the unified program system of the CMEA countries which will be used in the preparation of safety reports, the evaluation of safety hazards, the design of fuel charges, economical studies etc. A detailed description is also presented of multigroup transport calculations and of the preparation of input data for macrocalculations of the heterogeneous lattices of LWR's. (author)

  11. 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

  12. Control rod

    International Nuclear Information System (INIS)

    Igarashi, Takao; Yoshimoto, Yuichiro; Sugawara, Satoshi; Fukumoto, Takashi; Endo, Zen-ichiro; Saito, Shozo; Shinpo, Katsutoshi; Nishimura, Akira; Ozawa, Michihiro

    1988-01-01

    Purpose: To provide a sufficient shutdown margin upon reactor shutdown, prevent sheath deformation without decreasing neutron absorbents and prevent impact shocks exerted to structural materials. Constitution: The control rod of the present invention comprises a neutron absorption region, a sheath deformation means attached to the side wall and means for restricting and supporting axial movement of the neutron absorbent rod. Then, the amount of absorptive nuclei chained absorbents in the lower region is reduced than that in the upper region. In this way, effective neutron absorbing performance can be obtained relative to the neutron importance distribution during reactor shutdown. In addition, since the operationability is improved by reducing the weight of the control rod and the absorptive nuclei chained neutron abosrbers are used, mechanical nuclear life of the control rod can be increased. Thus, it is possible to prevent the outward deformation of the sheath, as well as prevent collision between the neutron absorber rod and the structural material on the side of inserting the control rod generated upon reactor scram by a simple structure. (Kamimura, M.)

  13. 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

  14. 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

  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. FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

    International Nuclear Information System (INIS)

    1996-09-01

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES ampersand H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ''Alternative Teams,'' chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S ampersand S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT's analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option

  17. FMDP reactor alternative summary report: Volume 4, Evolutionary LWR alternative

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    Significant quantities of weapons-usable fissile materials [primarily plutonium and highly enriched uranium (HEU)] have become surplus to national defense needs both in the United States and Russia. These stocks of fissile materials pose significant dangers to national and international security. The dangers exist not only in the potential proliferation of nuclear weapons but also in the potential for environmental, safety, and health (ES&H) consequences if surplus fissile materials are not properly managed. The purpose of this report is to provide schedule, cost, and technical information that will be used to support the Record of Process (ROD). Following the screening process, DOE/MD via its national laboratories initiated a more detailed analysis activity to further evaluate each of the ten plutonium disposition alternatives that survived the screening process. Three ``Alternative Teams,`` chartered by DOE and comprised of technical experts from across the DOE national laboratory complex, conducted these analyses. One team was chartered for each of the major disposition classes (borehole, immobilization, and reactors). During the last year and a half, the Fissile Materials Disposition Program (FMDP) Reactor Alternative Team (RxAT) has conducted extensive analyses of the cost, schedule, technical maturity, S&S, and other characteristics of reactor-based plutonium disposition. The results of the RxAT`s analyses of the existing LWR, CANDU, and partially complete LWR alternatives are documented in Volumes 1-3 of this report. This document (Volume 4) summarizes the results of these analyses for the ELWR-based plutonium disposition option.

  18. 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

  19. 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

  20. 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)

  1. Behavior of LWR fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Albrecht, H.; Bocek, M.; Erbacher, F.; Fiege, A.; Fischer, M.; Hagen, S.; Hofmann, P.; Holleck, H.; Karb, E.; Leistikow, S.; Melang, S.; Ondracek, G.; Thuemmler, F.; Wiehr, K.

    1977-01-01

    In the frame of the German reactor safety research program, the Kernforschungszentrum Karlsruhe is carrying out a comprehensive program on the behavior of LWR fuel elements under a variety of power cooling mismatch conditions in particular during loss-of-coolant accidents. The major objectives are to establish a detailed quantitative understanding of fuel rod failures mechanisms and their thresholds, to evaluate the safety margins of power reactor cores under accident conditions and to investigate the feedback of fuel rod failures on the efficiency of emergency core cooling systems. This detailed quantitative understanding is achieved through extensive basic and integral experiments and is incorporated in a fuel behavior code. On the basis of these results the design of power reactor fuel elements and of safety devices can be further improved. The results of investigations on the inelastic deformation (ballooning) behavior of Zircaloy 4 cladding at LOCA temperatures in oxidizing atmosphere are presented. Depending upon strain rate and temperature superplastic deformation behavior was observed. In the equation of state of Zry 4 the strain rate sensitivity index depends strongly upon strain and in the superplastic region upon sample anisotropy. Oxidation kinetics experiments with Zry-tubes at 900-1300 0 C showed that the Baker-Just correlation describes the reality quite conservative. Therefore a reduction of the amount of Zry oxidation can be assumed in the course of a LOCA. The external oxidation of Zry-cladding by steam as well as internal oxidation by the oxygen in oxide fuel and fission products (Cs, I, Te) have an influence on the strain and rupture behavior of Zry-cladding at LOCA temperatures. In out-of-pile and inpile experiments the mechanical and thermal behavior of fuel rods during the blowdown, the heatup and the reflood phases of a LOCA are investigated under representative and controlled thermohydraulic conditions. The task of the inpile experiments is

  2. 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

  3. 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).

  4. The light-water-reactor version of the Uranus integral fuel-rod code

    International Nuclear Information System (INIS)

    Moreno, A.; Lassmann, K.

    1977-01-01

    The LWR of the Uranus code, a digital computer programme for the thermal and mechanical analysis of fuel rods, is presented. Material properties are discussed and their effect on integral fuel rod behaviour elaborated via Uranus results for some carefully selected reference experiments. The numerical results do not represent post-irradiation analysis of in-pile experiments, they illustrate rather typical and diverse Uranus capabilities. The performance test shows that Uranus is reliable and efficient, thus the code is a most valuable tool in fuel fod analysis work. K. Lassmann developed the LWR version of the Uranus code, material properties were reviewed and supplied by A. Moreno. (author)

  5. Polycation induced actin bundles

    OpenAIRE

    Muhlrad, Andras; Grintsevich, Elena E.; Reisler, Emil

    2011-01-01

    Three polycations, polylysine, the polyamine spermine and the polycationic protein lysozyme were used to study the formation, structure, ionic strength sensitivity and dissociation of polycation-induced actin bundles. Bundles form fast, simultaneously with the polymerization of MgATP-G-actins, upon addition of polycations to solutions of actins at low ionic strength conditions. This indicates that nuclei and/or nascent filaments bundle due to attractive, electrostatic effect of polycations an...

  6. Polycation induced actin bundles.

    Science.gov (United States)

    Muhlrad, Andras; Grintsevich, Elena E; Reisler, Emil

    2011-04-01

    Three polycations, polylysine, the polyamine spermine and the polycationic protein lysozyme were used to study the formation, structure, ionic strength sensitivity and dissociation of polycation-induced actin bundles. Bundles form fast, simultaneously with the polymerization of MgATP-G-actins, upon the addition of polycations to solutions of actins at low ionic strength conditions. This indicates that nuclei and/or nascent filaments bundle due to attractive, electrostatic effect of polycations and the neutralization of repulsive interactions of negative charges on actin. The attractive forces between the filaments are strong, as shown by the low (in nanomolar range) critical concentration of their bundling at low ionic strength. These bundles are sensitive to ionic strength and disassemble partially in 100 mM NaCl, but both the dissociation and ionic strength sensitivity can be countered by higher polycation concentrations. Cys374 residues of actin monomers residing on neighboring filaments in the bundles can be cross-linked by the short span (5.4Å) MTS-1 (1,1-methanedyl bismethanethiosulfonate) cross-linker, which indicates a tight packing of filaments in the bundles. The interfilament cross-links, which connect monomers located on oppositely oriented filaments, prevent disassembly of bundles at high ionic strength. Cofilin and the polysaccharide polyanion heparin disassemble lysozyme induced actin bundles more effectively than the polylysine-induced bundles. The actin-lysozyme bundles are pathologically significant as both proteins are found in the pulmonary airways of cystic fibrosis patients. Their bundles contribute to the formation of viscous mucus, which is the main cause of breathing difficulties and eventual death in this disorder. Copyright © 2011 Elsevier B.V. All rights reserved.

  7. CONTROL ROD

    Science.gov (United States)

    Walker, D.E.; Matras, S.

    1963-04-30

    This patent shows a method of making a fuel or control rod for a nuclear reactor. Fuel or control material is placed within a tube and plugs of porous metal wool are inserted at both ends. The metal wool is then compacted and the tube compressed around it as by swaging, thereby making the plugs liquid- impervious but gas-pervious. (AEC)

  8. Fuel rods

    International Nuclear Information System (INIS)

    Hattori, Shinji; Kajiwara, Koichi.

    1980-01-01

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

  9. Rodding Surgery

    Science.gov (United States)

    ... Physical activity prior to surgery,  Length of the operation; anesthesia issues,  Reason for the choice of rod,  Time in the hospital,  Length of recovery time at home,  Pain management including control of muscle spasms,  The rehabilitation plan. ...

  10. Control rods

    International Nuclear Information System (INIS)

    Koga, Isao; Masuoka, Ryuzo.

    1979-01-01

    Purpose: To prevent fuel element failures during power conditioning by removing liquid absorbents in poison tubes of control rods in a fast power up step and extracting control rods to slightly increase power in a medium power up step. Constitution: A plurality of poison tubes are disposed in a coaxial or plate-like arrangement and divided into a region capable of compensating the reactivity from the initial state at low temperature to 40% power operation and a region capable of compensating the reactivity in the power up above 40% power operation. Soluble poisons are used as absorbers in the poison tubes corresponding to above 40% power operation and they are adapted to be removed independently from the driving of control rods. The poison tubes filled with the soluble absorbers are responsible for the changes in the reactivity from the initial state at low temperature to the medium power region and the reactivity control is conducted by the elimination of liquid absorbers from the poison tubes. In the succeeding slight power up region above the medium power, power up is proceeding by extracting the control rods having remaining poison tubes filled with solid or liquid absorbers. (Seki, T.)

  11. Outline of Swedish activities on LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Grounes, M [Studsvik Nuclear, Nykoeping (Sweden); Roennberg, G [OKG AB (Sweden)

    1997-12-01

    The presentation outlines the Swedish activities on LWR fuel and considers the following issues: electricity production; performance of operating nuclear power plants; nuclear fuel cycle and waste management; research and development in nuclear field. 4 refs, 4 tabs.

  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. 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

  14. Proceedings of the 2007 LWR Fuel Performance Meeting / TopFuel 2007 'Zero by 2010'

    International Nuclear Information System (INIS)

    2007-01-01

    ANS, ENS, AESJ and KNS are jointly organizing the 2007 International LWR Fuel Performance Meeting following the successful ENS TopFuel meeting held during 22-26 October, 2006 in Salamaca, Spain. Merging three premier nuclear fuel design and performance meetings: the ANS LWR Fuel Performance Meeting, the ENS TopFuel and Asian Water Reactor Fuel Performance Meeting (WRFPM) created this international meeting. The meeting will be held annually on a tri-annual rotational basis in USA, Asia, and Europe. The technical scope of the meeting includes all aspects of nuclear fuel from fuel rod to core design as well as performance experience in commercial and test reactors. The meeting excludes front end and back end fuel issues, however, it covers all front and/or back issues that impact fuel designs and performance

  15. Principal noncommutative torus bundles

    DEFF Research Database (Denmark)

    Echterhoff, Siegfried; Nest, Ryszard; Oyono-Oyono, Herve

    2008-01-01

    of bivariant K-theory (denoted RKK-theory) due to Kasparov. Using earlier results of Echterhoff and Williams, we shall give a complete classification of principal non-commutative torus bundles up to equivariant Morita equivalence. We then study these bundles as topological fibrations (forgetting the group...

  16. Validation of KENOREST with LWR-PROTEUS phase II samples

    Energy Technology Data Exchange (ETDEWEB)

    Wagner, M.; Kilger, R.; Pautz, A.; Zwermann, W. [GRS, Garching (Germany); Grimm, P.; Vasiliev, A.; Ferroukhi, H. [Paul Scherrer Institut, Villigen (Switzerland)

    2012-11-01

    In order to broaden the validation basis of the reactivity and nuclide inventory code KENOREST two samples of the LWR-PROTEUS phase II program have been calculated and compared to the experimental results. In general most nuclides are reproduced very well and agree within about ten percent with the experiment. Some already known problems, the overprediction of metallic fission products and the underprediction of the higher curium isotopes, have been confirmed. One of the largest uncertainties in the calculation was the burnup of the samples due to differences between a core simulation of the fuel vendor and the burnup determined from the measured values of the burnup indicator Nd-148. Two different models taking into account the environment for a peripheral fuel rod have been studied. The more detailed model included the three direct neighbor fuel assemblies depleted along with the fuel rod of interest. The influence on the results has been found to be very small. Compared to the uncertainties from the burnup, this effect can be considered negligible. The reason for the low influence was basically that the spectrum did not get considerably harder with increasing burnup beyond about 20GWd/tHM. Since the sample reached burnups far beyond that value, an effect could not be seen. In the near future an update of the used libraries is planned and it will be very interesting to study the effect on the results, especially for Curium. (orig.)

  17. 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.

  18. Control rod

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Inoue, Kotaro.

    1979-01-01

    Purpose: To flatten the power distribution in the reactor core without impairing neutron economy by disposing pins containing elements of lower atomic number in the central region of a shroud and loading pins containing depleted uranium in the periphery region thereof. Constitution: The shroud has a layer of pins containing depleted uranium in the peripheral region and a layer of pins containing elements of lower atomic number such as beryllium in the central region. Heat removal from those pins containing depleted uranium and elements of lower atomic number (neutron moderator) is effected by sodium flow outside of the cladding material. The control rod operation is conducted by inserting or extracting the central portion (pins containing elements of lower atomic number such as beryllium) inside of the stainless pipe. Upon extraction of the control rod, the moderator in the central region is removed whereby high speed neutrons are no more deccelerated and the absorption rate to the depleted uranium is decreased. This can flatten the power distribution in the reactore core with the disposition of a plurality of control rods at a better neutron economy as compared with the use of neutron absorber such as boron. (Seki, T.)

  19. Control rod

    International Nuclear Information System (INIS)

    Fukumoto, Takashi; Hirakawa, Hiromasa; Kawashima, Norio; Goto, Yasuyuki.

    1994-01-01

    Neutron absorbers are contained in a tubular member comprising, integrally a tubular portion and four corners disposed at the outer circumference of the tubular portion at every 90deg, to provide a neutron absorbing tube. A plurality of neutron absorbing tubes are arranged in parallel in the lateral direction, and adjacent corners are joined, into a blade to constitute a control rod. Such a control rod has a great structural strength, simple in the structure and relatively light in weight and can contain a great amount of neutron absorbers. Upon formation of the control rod by arranging the blades in a cross-like shape, at least a portion thereof is constituted with short neutron absorbing tubes shorter than the entire length of the blade, and gaps are formed at positions in adjacent in the axial direction. With such a constitution, there is no worry that a wing end of the blade collides against or be abraded with a fuel channel box or a fuel support. Even if fuel channels are vibrated upon scram of the reactor, such as occurrence of earthquakes, it can be inserted to the reactor easily. (N.H.)

  20. Performance and reliability of LWR fuel

    International Nuclear Information System (INIS)

    Bairiot, H.; Deramaix, P.; Vandenberg, C.

    1977-01-01

    The main requirements for fuel reloads are: good reliability, minimum fuel cycle costs and flexibility of operation. Fulfilling these goals requires a background of experience. The approach to the acquisition of this experience in the particular case of BN has included over the last 15 years a proper development and cross-checking of the design methods and criteria, a continuous updating of the drawings and specifications and the qualification of adequate fabrication plants. This approach can best be outlined on the basis of the gradual implementation of the modern features of the LWR fuel. The first fuel clad with stainless steel was loaded in the BR 3 (11 MWe) in 1969 and later on (since 1974) in the SENA plant (310 MWe). Similarly, Zircaloy 4 cladding was first introduced in a reactor reload in 1969 as autoclaved cladding and later on (in 1971) the autoclaving was suppressed for the further reloads. Zircaloy 2 was loaded in DODEWAARD (51.5 MWe) in 1970. The first demonstration assembly in a PWR was a Pu-island assembly loaded in the BR 3 in 1963. It was followed by an all-Pu assembly in the same reactor in 1965 and by the loading of Pu fuels in four prototype assemblies in GARIGLIANO (160 MWe) in 1968. A full reload incorporating Pu fuel has been experienced by the supply of fuel for GARIGLIANO (BOL: 1975) and for BR 3 (BOL: 1972 and 1976). While in the early sixties the brazed design was still being utilized, the first assembly incorporating grids with springs was introduced in BR 3 in 1963. The first Inconel grids were loaded in the same reactor in 1969 and the first Zircaloy grids in 1972 (the first Zr grid has been loaded in a BWR in 1973). The experience covered successively the shrouded design (BOL: 1963), the shroudless design (BOL: 1969), a BWR assembly (BOL: 1971), a typical RCC assembly first with large diameter fuel rods (1972) and later on with small diameter fuel rods (1974). The experience on the reactivity control covered successively diluted

  1. Preliminary reactor physics calculations for Exxon LWR fuel testing in the power burst facility

    International Nuclear Information System (INIS)

    Olson, W.O.; Nigg, D.W.

    1981-05-01

    The PFB reactor is being considered as an irradiation facility to test LWR fuel rods for Exxon Nuclear Company. Requested test conditions are 18 kW/ft axial peak steady state power in 2.5% initial enrichment, 20,000 MWd/Tu exposed rods. Multigroup transport theory calculations (S/sub n/ and Monte Carlo) showed that this was unattainable in the standard PBF test loop. Thus, a flux multiplier was developed in the form of a Zr-2-clad 0.15-inch thick cylindrical shell of 35% enriched, 88% T.D. UO 2 replacing the flow divider, surrounding the rod within the in-pile tube in PFB. With this flux multiplier installed and assuming an average water density of 0.86 g/cm 3 within the test loop, a Figure of Merit (FOM) for a single-rod test assembly of 0.86 kW/ft-MW +- 5% (at 95% confidence level) was calculated. This FOM is the axial peak linear test rod power per megawatt of reactor power. A reactor power of about 21 megawatts will therefore be required to supply the requested linear test rod axial peak heating rate of 18 kW/ft

  2. 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

  3. 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

  4. Sucker rods

    Energy Technology Data Exchange (ETDEWEB)

    Rylov, B M; Kostur, I N; Shcheigiy, B I; Sukhanov, V S

    1983-01-01

    As an addendum to A.s. USSR patent No 769087, this particular sucker rod utilizes a differential piston spring that has been attached outside the body of the auxiliary pump. The pump cylinder is attached to the intake line of the main pump. The lower part of the auxiliary pump is equipped with vertical slits, while the differential piston is equipped with a perforated pusher and support under the spring; it can also be shifted as necessary with respect to the vertical slits.

  5. Matpro--version 10: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    International Nuclear Information System (INIS)

    Reymann, G.A.

    1978-02-01

    The materials properties correlations and computer subcodes (MATPRO--Version 10) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory are described. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures

  6. MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    International Nuclear Information System (INIS)

    Hagrman, D.L.; Reymann, G.A.

    1979-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures

  7. MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

    Energy Technology Data Exchange (ETDEWEB)

    Hagrman, D.L.; Reymann, G.A. (comps.)

    1979-02-01

    This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.

  8. 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.

  9. 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

  10. Fatigue management considering LWR coolant environments

    International Nuclear Information System (INIS)

    Park, Heung Bae; Jin, Tae eun

    2000-01-01

    Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)

  11. 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

  12. Fuel rods

    International Nuclear Information System (INIS)

    Adachi, Hajime; Ueda, Makoto

    1985-01-01

    Purpose: To provide a structure capable of measuring, in a non-destructive manner, the releasing amount of nuclear gaseous fission products from spent fuels easily and at a high accuracy. Constitution: In order to confirm the integrity and the design feasibility of a nuclear fuel rod, it is important to accurately determine the amount of gaseous nuclear fission products released from nuclear pellets. In a structure where a plurality of fuel pellets are charged in a fuel cladding tube and retained by an inconel spring, a hollow and no-sealed type spacer tube made of zirconium or the alloy thereof, for example, not containing iron, cobalt, nickel or manganese is formed between the spring and the upper end plug. In the fuel rod of such a structure, by disposing a gamma ray collimator and a gamma ray detector on the extension of the spacer pipe, the gamma rays from the gaseous nuclear fission products accumulated in the spacer pipe can be detected while avoiding the interference with the induction radioactivity from inconel. (Kamimura, M.)

  13. Characterization and chemistry of fission products released from LWR fuel under accident conditions

    International Nuclear Information System (INIS)

    Norwood, K.S.; Collins, J.L.; Osborne, M.F.; Lorenz, R.A.; Wichner, R.P.

    1984-01-01

    Segments from commercial LWR fuel rods have been tested at temperatures between 1400 and 2000 0 C in a flowing steam-helium atmosphere to simulate severe accident conditions. The primary goals of the tests were to determine the rate of fission product release and to characterize the chemical behavior. This paper is concerned primarily with the identification and chemical behavior of the released fission products with emphasis on antimony, cesium, iodine, and silver. The iodine appeared to behave primarily as cesium iodide and the antimony and silver as elements, while cesium behavior was much more complex. 17 refs., 7 figs., 1 tab

  14. 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

  15. 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.)

  16. Information on the evolution of severe LWR fuel element damage obtained in the CORA program

    International Nuclear Information System (INIS)

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

    1992-01-01

    In the CORA program a series of out-of-pile experiments on LWR severe accidental situations is being performed, in which test bundles of LWR typical components and arrangements (PWR, BWR) are exposed to temperature transients up to about 2400deg C under flowing steam. The individual features of the facility, the test conduct, and the evaluation will be presented. In the frame of the international cooperation in severe fuel damage (SFD) programs the CORA tests are contributing confirmatory and complementary informations to the results from the limited number of in-pile tests. The identification of basic phenomena of the fuel element destruction, observed as a function of temperature, is supported by separate-effects test results. Most important mechanisms are the steam oxidation of the Zircaloy cladding, which determines the temperature escalation, the chemical interaction between UO 2 fuel and cladding, which dominates fuel liquefaction, relocation and resulting blockage formation, as well as chemical interactions with Inconel spacer grids and absorber units ((Ag, In, Cd) alloy or B 4 C), which are leading to extensive low-temperature melt formation around 1200deg C. Interrelations between those basic phenomena, resulting for example in cladding deformation ('flowering') and the dramatic hydrogen formation in response to the fast cooling of a hot bundle by cold water ('quenching') are determining the evolution paths of fuel element destruction, which are to be identified. (orig.)

  17. 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

  18. 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.

  19. 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)

  20. 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.)

  1. LWR control assembly designs: A historical perspective

    International Nuclear Information System (INIS)

    Kennard, M.W.; Harbottle, J.E.

    2000-01-01

    Control rod designs and materials have evolved in response to performance problems in both PWRs and BWRs. Irradiation-assisted stress corrosion cracking (IASCC) due to absorber swelling has primarily affected BWR control rods with B 4 C absorbers, but has also occurred in PWRs with Ag-In-Cd absorbers. The primary problems for some designs of PWR control rods have been wear of the rodlets against upper internal components and swelling with tip wear and cracking. Competition amongst vendors for supplying control rod reloads has also resulted in design improvements. This paper provides an historical review of PWR and BWR control rod designs, their problems and remedies. (author)

  2. 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

  3. Fuel rods

    International Nuclear Information System (INIS)

    Fukushima, Kimichika.

    1984-01-01

    Purpose: To reduce the size of the reactor core upper mechanisms and the reactor container, as well as decrease the nuclear power plant construction costs in reactors using liquid metals as the coolants. Constitution: Isotope capturing devices comprising a plurality of pipes are disposed to the gas plenum portion of a nuclear fuel rod main body at the most downstream end in the flowing direction of the coolants. Each of the capturing devices is made of nickel, nickel alloys, stainless steel applied with nickel plating on the surface, nickel alloys applied with nickel plating on the surface or the like. Thus, radioactive nuclides incorporated in the coolants are surely captured by the capturing devices disposed at the most downstream end of the nuclear fuel main body as the coolants flow along the nuclear fuel main body. Accordingly, since discharging of radioactive nuclides to the intermediate fuel exchange system can be prevented, the maintenance or reparing work for the system can be facilitated. (Moriyama, K.)

  4. Control rods

    International Nuclear Information System (INIS)

    Hirukawa, Koji.

    1979-01-01

    Purpose: To ensure the fuel safety by constituting a control rod with a plurality of poison bodies suspended in a cross-like section and shorter length poison bodies made movable and engageable in the gap between each of the above poison bodies thereby maintaining the function of the shorter length poison constant. Constitution: Cross-like supports are secured to the upper and lower parts of a driving shaft journaled in a sheath and poison bodies composed of neutron absorber poisons of a large thermal neutron absorption cross section and neutron absorber poison tubes for containing them are suspended from the supports. A movable cross-like support is mounted slidably at its base to the lower part of the driving shaft and poison bodies shorter than the above poison bodies and composed of neutron absorber poisons having a greater absorption cross section at the neutron energy region higher than thermal neutron region and neutron poison tubes for containing them are suspended to the movable support at the position capable of engaging in the gap between each of the poison bodies. (Kawakami, Y.)

  5. RETRANS, Reactivity Transients in LWR

    International Nuclear Information System (INIS)

    Kamelander, G.

    1989-01-01

    1 - Description of program or function: RETRANS is appropriate to calculate power excursions in light water reactors initiated by reactivity insertions due to withdrawal of control elements. As in the code TWIGL, the neutron physics model is based on the time-dependent two-group neutron diffusion equations. The equation of state of the coolant is approximated by a table built into the code. RETRANS solves the heat conduction equation and calculates the heat transfer coefficient for representative fuel rods at each time-step. 2 - Method of solution: The time-dependent neutron diffusion equations are modified by an exponential transformation and solved by means of a finite difference method. There is an option accelerating the inner iterations of the difference scheme by a coarse-mesh re-balancing method. The heat balance equations of the thermo- hydraulic model are discretized and converted into a tri-diagonal system of linear equations which is solved recursively. 3 - Restrictions on the complexity of the problem: r-z-geometry, one- phase-flow

  6. A comparison of thermal algorithms of fuel rod performance code systems

    International Nuclear Information System (INIS)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C.

    2003-11-01

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance

  7. A comparison of thermal algorithms of fuel rod performance code systems

    Energy Technology Data Exchange (ETDEWEB)

    Park, C. J.; Park, J. H.; Kang, K. H.; Ryu, H. J.; Moon, J. S.; Jeong, I. H.; Lee, C. Y.; Song, K. C

    2003-11-01

    The goal of the fuel rod performance is to identify the robustness of a fuel rod with cladding material. Computer simulation of the fuel rod performance becomes one of important parts to designed and evaluate new nuclear fuels and claddings. To construct a computing code system for the fuel rod performance, several algorithms of the existing fuel rod performance code systems are compared and are summarized as a preliminary work. Among several code systems, FRAPCON, and FEMAXI for LWR, ELESTRES for CANDU reactor, and LIFE for fast reactor are reviewed. Thermal algorithms of the above codes are investigated including methodologies and subroutines. This work will be utilized to construct a computing code system for dry process fuel rod performance.

  8. Standard-model bundles

    CERN Document Server

    Donagi, Ron; Pantev, Tony; Waldram, Dan; Donagi, Ron; Ovrut, Burt; Pantev, Tony; Waldram, Dan

    2002-01-01

    We describe a family of genus one fibered Calabi-Yau threefolds with fundamental group ${\\mathbb Z}/2$. On each Calabi-Yau $Z$ in the family we exhibit a positive dimensional family of Mumford stable bundles whose symmetry group is the Standard Model group $SU(3)\\times SU(2)\\times U(1)$ and which have $c_{3} = 6$. We also show that for each bundle $V$ in our family, $c_{2}(Z) - c_{2}(V)$ is the class of an effective curve on $Z$. These conditions ensure that $Z$ and $V$ can be used for a phenomenologically relevant compactification of Heterotic M-theory.

  9. 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

  10. The Light-Water-Reactor Version of the URANUS Integral fuel-rod code

    Energy Technology Data Exchange (ETDEWEB)

    Labmann, K; Moreno, A

    1977-07-01

    The LWR version of the URANUS code, a digital computer programme for the thermal and mechanical analysis of fuel rods, is presented. Material properties are discussed and their effect on integral fuel rod behaviour elaborated via URANUS results for some carefully selected reference experiments. The numerical results do not represent post-irradiation analyses of in-pile experiments, they illustrate rather typical and diverse URANUS capabilities. The performance test shows that URANUS is reliable and efficient, thus the code is a most valuable tool in fuel rod analysis work. K. LaBmann developed the LWR version of the URANUS code, material properties were reviewed and supplied by A. Moreno. (Author) 41 refs.

  11. Assessment of dry storage performance of spent LWR fuel assemblies with increasing burnup

    International Nuclear Information System (INIS)

    Peehs, M.; Garzarolli, F.; Goll, W.

    1999-01-01

    Although the safety of a dry long-term spent fuel store is scarcely influenced if a few fuel rods start to leak during extended storage - since all confinement systems are designed to retain gaseous activity safely - it is a very conservative safety goal to avoid the occurrence of systematic rod defects. To assess the extended storage performance of a spent fuel assembly (FA), the experience can be collated into 3 storage modes: I - fast rate of temperature decrease δ max ≥ δ ≥ 300 deg. C, II - medium rate of decrease for the fuel rod dry storage temperature 300 deg. C > δ ≥ 200 deg. C, III - slow to negligible rate of temperature decrease for δ 2 -fuel are practically immobile during storage. Consequently all fission-product-driven defect mechanisms will not take place. The leading defect mechanism - also for fuel rods with increased burnup - remains creep due to the hoop strain resulting from the fuel rod internal fission gas pressure. Limiting the creep to its primary and secondary stages prevents fuel rod degradation. The allowable uniform strain of the cladding is 1 - 2%. Calculations were performed to predict the dry storage performance of fuel assemblies with a burnup ≤ 55 GW · d/tHM based on the fuel assemblies end of life (EOL)-data and on a representative curve T = f(t). The maximum allowable hot spot temperature of a fuel rod in the CASTOR V cask was between 348 deg. C (U FA) and 358 deg. C (MOX FA). The highest hoop strain predicted after 40 years of storage is 0.77% proving that spent LWR fuel dry storage is safe. (author)

  12. 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)

  13. Studies on the effects of blockage upon LWR emergency core cooling systems

    International Nuclear Information System (INIS)

    Fairbairn, S.A.; Piggott, B.D.G.

    1985-01-01

    Ballooning of the zircaloy cladding of PWR fuel pins could occur during certain postulated Loss of Coolant Accidents. This report describes experimental data obtained in a 44-rod bundle with and without a localized coplanar blockage under conditions relevant to the reflood phase of a LOCA. The aim of the work is to provide a data base for modelling dispersed flow heat transfer around a local blockage. This work concentrates on the thermohydraulic aspects of the ballooning problem by use of pre-formed balloon shapes attached to the rods of an electrically heated rod bundle. The various thermohydraulic effects are investigated separately, as far as possible, in a unique series of tests of increasing complexity proceeding from single to two phase conditions as follows: isothermal air flow tests, used to infer the single phase mass flow distribution; steady state steam flow tests, used to quantify single phase heat transfer; steam and droplet tests, in which a dispersed flow of well specified inlet conditions is created by injecting water droplets into the subchannel centres between the rods with a co-current steam flow; and finally, conventional reflood tests. The first part makes an extensive presentation of all the data obtained for an undistorted bundle and a bundle containing a centrally placed 4x4 array of balloon shapes (approximately 50 mm long, solid) which create a 90% subchannel blockage at their centre elevations. In part 2 tests on two blockage shapes each producing 90% subchannel blockage are described. The first shape is composed of thick walled sleeves (1.0 to 2.5 mm) and the second of sleeves with a more realistic thermal capacity being only about 0.3 mm thick. 48 refs., 335 figs.

  14. Irradiated fuel bundle counter

    International Nuclear Information System (INIS)

    Campbell, J.W.; Todd, J.L.

    1975-01-01

    The design of a prototype safeguards instrument for determining the number of irradiated fuel assemblies leaving an on-power refueled reactor is described. Design details include radiation detection techniques, data processing and display, unattended operation capabilities and data security methods. Development and operating history of the bundle counter is reported. (U.S.)

  15. Irradiated fuel bundle counter

    International Nuclear Information System (INIS)

    Campbell, J.W.; Todd, J.L.

    1975-01-01

    The design of a prototype safeguards instrument for determining the number of irradiated fuel assemblies leaving an on-power refueled reactor is described. Design details include radiation detection techniques, data processing and display, unattended operation capabilities and data security methods. Development and operating history of the bundle counter is reported

  16. ALUMINUM BOX BUNDLING PRESS

    Directory of Open Access Journals (Sweden)

    Iosif DUMITRESCU

    2015-05-01

    Full Text Available In municipal solid waste, aluminum is the main nonferrous metal, approximately 80- 85% of the total nonferrous metals. The income per ton gained from aluminum recuperation is 20 times higher than from glass, steel boxes or paper recuperation. The object of this paper is the design of a 300 kN press for aluminum box bundling.

  17. Kernel bundle EPDiff

    DEFF Research Database (Denmark)

    Sommer, Stefan Horst; Lauze, Francois Bernard; Nielsen, Mads

    2011-01-01

    In the LDDMM framework, optimal warps for image registration are found as end-points of critical paths for an energy functional, and the EPDiff equations describe the evolution along such paths. The Large Deformation Diffeomorphic Kernel Bundle Mapping (LDDKBM) extension of LDDMM allows scale space...

  18. SSYST-1. A computer code system to analyse the fuel rod behaviour during a loss of coolant accident

    International Nuclear Information System (INIS)

    Gulden, W.

    1977-08-01

    The modules of the SSYST program system allow the detailed analysis of an LWR fuel rod in the course of a postulated loss-of-coolant accident. They provide a tool for considering the interaction between the heat conduction in the fuel rod, heat transfer in the gap, fuel and cladding tube deformation, pressure in the coolant, as well as thermal and fluid dynamics in the cooling channel and for calculating the time and location of ballooning and rod failure, respectively. They can be used both to precalculate the behaviour of fuel rods during LWR accidents and in support of the design of experiments. Depending on the problem to be solved, the individual modules can be easily combined. (orig.) [de

  19. Control rod drives

    International Nuclear Information System (INIS)

    Nakamura, Akira.

    1984-01-01

    Purpose: To enable to monitor the coupling state between a control rod and a control rod drive. Constitution: After the completion of a control rod withdrawal, a coolant pressure is applied to a control rod drive being adjusted so as to raise only the control rod drive and, in a case where the coupling between the control rod drive and the control rod is detached, the former is elevated till it contacts the control rod and then stopped. The actual stopping position is detected by an actual position detection circuit and compared with a predetermined position stored in a predetermined position detection circuit. If both of the positions are not aligned with each other, it is judged by a judging circuit that the control rod and the control rod drives are not combined. (Sekiya, K.)

  20. 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

  1. 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

  2. Development of training simulator for LWR

    International Nuclear Information System (INIS)

    Sureshbabu, R.M.

    2009-01-01

    A full-scope training simulator was developed for a light water reactor (LWR). This paper describes how the development evolved from a desktop simulator to the full-scope training simulator. It also describes the architecture and features of the simulator including the large number of failures that it simulates. The paper also explains the three-level validation tests that were used to qualify the training simulator. (author)

  3. 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)

  4. 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)

  5. 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

  6. 'CANDLE' burnup regime after LWR regime

    International Nuclear Information System (INIS)

    Sekimoto, Hiroshi; Nagata, Akito

    2008-01-01

    CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment. If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth. The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium. (author)

  7. Modeling the economic consequences of LWR accidents

    International Nuclear Information System (INIS)

    Burke, R.P.; Aldrich, D.C.; Rasmussen, N.C.

    1984-01-01

    Models to be used for analyses of economic risks from events which may occur during LWR plant operation are developed in this study. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The models can be used by both the nuclear power industry and regulatory agencies in cost-benefit analyses for decisionmaking purposes. The newly developed economic consequence models are applied in an example to estimate the economic risks from operation of the Surry Unit 2 plant. The analyses indicate that economic risks from US LWR operation, in contrast to public health risks, are dominated by relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for the Surry site. The implications of these conclusions for nuclear power plant operation and regulation are discussed

  8. Technical report on LWR design decision methodology. Phase I

    International Nuclear Information System (INIS)

    1980-03-01

    Energy Incorporated (EI) was selected by Sandia Laboratories to develop and test on LWR design decision methodology. Contract Number 42-4229 provided funding for Phase I of this work. This technical report on LWR design decision methodology documents the activities performed under that contract. Phase I was a short-term effort to thoroughly review the curret LWR design decision process to assure complete understanding of current practices and to establish a well defined interface for development of initial quantitative design guidelines

  9. 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

  10. 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

  11. 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

  12. Status of LWR fuel design and future usage of JENDL

    International Nuclear Information System (INIS)

    Ito, Takuya

    2008-01-01

    For all conventional LWR fuel design codes of LWR fuel manufactures in Japan, the cross section library are based on the ENDF/B. Recently we can see several movements for the utilization of JENDL library for the LWR fuel design. The latest version of NEUPHYS cross section library is based on the JENDL-3.2. To accelerate this movement of JENDL utilization in LWR fuel design, it is necessary to prepare a high quality JENDL document, systematic validation of JENDL and to appeal them abroad effectively. (author)

  13. Implementation of static generalized perturbation theory for LWR design applications

    International Nuclear Information System (INIS)

    Byron, R.F.; White, J.R.

    1987-01-01

    A generalized perturbation theory (GPT) formulation is developed for application to light water reactor (LWR) design. The extensions made to standard generalized perturbation theory are the treatment of thermal-hydraulic and fission product poisoning feedbacks, and criticality reset. This formulation has been implemented into a standard LWR design code. The method is verified by comparing direct calculations with GPT calculations. Data are presented showing that feedback effects need to be considered when using GPT for LWR problems. Some specific potential applications of this theory to the field of LWR design are discussed

  14. Control rod assembly

    International Nuclear Information System (INIS)

    Takahashi, Akio.

    1982-01-01

    Purpose: To enable reliable insertion and drops of control rods, as well as insure a sufficient flow rate of coolants flowing through the control rods for attaining satisfactory cooling thereof to enable relexation of thermal stress resulted to rectifying mechanisms or the likes. Constitution: To the outer circumference of a control rod contained vertically movably within a control rod guide tube, resistive members are retractably provided in such a way as to project to close the gap between outer circumference of the control rod and the inner surface of the control rod guide tube upon engagement of a gripper of control rod drives, and retract upon release of the engagement of the gripper. Thus, since the resistive members project to provide a greater resistance to the coolants flowing between them and the control rod guide tube in the normal operation where the gripper is engaged to drive the control rod by the control rod drives, a major part of the coolant flowing into the control rod guide tube flows into the control rod. This enables to cool the control rod effectively and make the temperature distribution uniform for the coolant flowing from the upper end of the control rod guide tube to thereby attain the relaxation of the thermal stress resulted in the rectifying mechanisms or the likes. (Moriyama, K.)

  15. Control rod displacement

    International Nuclear Information System (INIS)

    Nakazato, S.

    1987-01-01

    This patent describes a nuclear reactor including a core, cylindrical control rods, a single support means supporting the control rods from their upper ends in spaced apart positions and movable for displacing the control rods in their longitudinal direction between a first end position in which the control rods are fully inserted into the core and a second end position in which the control rods are retracted from the core, and guide means contacting discrete regions of the outer surface of each control rod at least when the control rods are in the vicinity of the second end position. The control rods are supported by the support means for longitudinal movement without rotation into and out of the core relative to the guide means to thereby cause the outer surface of the control rods to experience wear as a result of sliding contact with the guide means. The support means are so arranged with respect to the core and the guide means that it is incapable of rotation relative to the guide means. The improvement comprises displacement means being operatively coupled to a respective one of the control rods for periodically rotating the control rod in a single angular direction through an angle selected to change the locations on the outer surfaces of the control rods at which the control rods are contacted by the guide means during subsequent longitudinal movement of the control rods

  16. 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

  17. 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

  18. Right bundle branch block

    DEFF Research Database (Denmark)

    Bussink, Barbara E; Holst, Anders Gaarsdal; Jespersen, Lasse

    2013-01-01

    AimsTo determine the prevalence, predictors of newly acquired, and the prognostic value of right bundle branch block (RBBB) and incomplete RBBB (IRBBB) on a resting 12-lead electrocardiogram in men and women from the general population.Methods and resultsWe followed 18 441 participants included...... in the Copenhagen City Heart Study examined in 1976-2003 free from previous myocardial infarction (MI), chronic heart failure, and left bundle branch block through registry linkage until 2009 for all-cause mortality and cardiovascular outcomes. The prevalence of RBBB/IRBBB was higher in men (1.4%/4.7% in men vs. 0.......5%/2.3% in women, P block was associated with significantly...

  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. Bundling harvester; Nippukorjausharvesteri

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, K. [Eko-Log Oy, Kuopio (Finland)

    1996-12-31

    The staring point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automatizing of the harvester, and automatized loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilization of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilized without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilization of wood-energy

  1. Bundling harvester; Nippukorjausharvesteri

    Energy Technology Data Exchange (ETDEWEB)

    Koponen, K [Eko-Log Oy, Kuopio (Finland)

    1997-12-31

    The staring point of the project was to design and construct, by taking the silvicultural point of view into account, a harvesting and processing system especially for energy-wood, containing manually driven bundling harvester, automatizing of the harvester, and automatized loading. The equipment forms an ideal method for entrepreneur`s-line harvesting. The target is to apply the system also for owner`s-line harvesting. The profitability of the system promotes the utilization of the system in both cases. The objectives of the project were: to construct a test equipment and prototypes for all the project stages, to carry out terrain and strain tests in order to examine the usability and durability, as well as the capacity of the machine, to test the applicability of the Eko-Log system in simultaneous harvesting of energy and pulp woods, and to start the marketing and manufacturing of the products. The basic problems of the construction of the bundling harvester have been solved using terrain-tests. The prototype machine has been shown to be operable. Loading of the bundles to form sufficiently economically transportable loads has been studied, and simultaneously, the branch-biomass has been tried to be utilized without loosing the profitability of transportation. The results have been promising, and will promote the profitable utilization of wood-energy

  2. Flooding of a large, passive, pressure-tube LWR

    Energy Technology Data Exchange (ETDEWEB)

    Hejzlar, P.; Todreas, N.E.; Driscoll, M.J. [Massachusetts Institute of Technology, Cambridge, MA (United States)

    1995-09-01

    A reactor concept has been developed which can survive LOCA without scram and without replenishing primary coolant inventory. The proposed concept is a pressure tube type reactor similar to CANDU reactors, but differing in three key aspects: (1) a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles, (2) the heavy water coolant in the pressure tubes is replaced by light water, and (3) the calandria tank contains a low pressure gas instead of heavy water moderator. The gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents, it allows passive calandria flooding. This paper describes the thermal hydraulic characteristics of the gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube LWR concept. The flooding of the top row of fuel channels must be accomplished fast enough so that none of the critical components of the fuel channel exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. Two other considerations are important. The thermal shock experienced by the calandria and pressure tubes has been evaluated and shown to be within acceptable bounds. Finally, although complete flooding renders the reactor deeply subcritical, various steam/water densities can be hypothesized to be present during the flooding process which could cause reactivity to increase from the initially voided calandria case. One such hypothesis which leads to the maximum possible density of the steam/water mixture in the still unflooded calandria space is entrainment from the free surface. It is shown that the steam/water mixture density yielding the maximum reactivity peak cannot be achieved by entrainment because it exceeds thermohydraulically attainable densities of steam/water by an order of magnitude.

  3. Control rod drive mechanism

    International Nuclear Information System (INIS)

    Nakamura, Akira.

    1981-01-01

    Purpose: To ensure the scram operation of a control rod by the reliable detection for the position of control rods. Constitution: A permanent magnet is provided to the lower portion of a connecting rod in engagement with a control rod and a tube having a plurality of lead switches arranged axially therein in a predetermined pitch is disposed outside of the control rod drives. When the control rod moves upwardly in the scram operation, the lead switches are closed successively upon passage of the permanent magnet to operate the electrical circuit provided by way of each of the lead switches. Thus, the position for the control rod during the scram can reliably be determined and the scram characteristic of the control rod can be recognized. (Furukawa, Y.)

  4. 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.

  5. 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)

  6. 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

  7. The Atiyah bundle and connections on a principal bundle

    Indian Academy of Sciences (India)

    be the fiber bundle constructed as in (1.1) for the universal principal G-bundle. In a work in progress, we hope to show that the universal G-connection can be realized as a fiber bundle over C(EG). Turning this ... a G-invariant vector field on EG|U . In other words, we get a bijective linear map between. A(EG)(U) (the space of ...

  8. Creep damage in zircaloy-4 at LWR temperatures

    International Nuclear Information System (INIS)

    Keusseyan, R.L.; Hu, C.P.; Li, C.Y.

    1978-08-01

    The observation of creep damage in the form of grain boundary cavitation in Zircaloy-4 in the temperature range of interest to Light Water Reactor (LWR) applications is reported. The observed damage is shown to reduce the ductility of Zircaloy-4 in a tensile test at LWR temperatures

  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. 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. Is it the end of history for LWR safety?

    International Nuclear Information System (INIS)

    Sehgal, Bal Raj

    2004-01-01

    In this essay a parallel is drawn between the struggle for recognition, which is argued by Fukuyama as the 'motor' of human history and that waged by the LWR safety for the public to recognize the LWR plants as a source of safe nuclear power. The end of history for the ''human struggle for recognition'' as the capitalistic liberal democracy is equated with the ''end of history'' for the LWR safety to provide assurance to the public of termination of a severe accident it ever would occur. It is suggested that we are near ''the end of history'' of the LWR safety for the new-design LWR plants but fall short for the presently-installed plants. The essay bases these suggestions on an examination of the history of nuclear power development in U.S.A., but also considering the more recent regulatory and public acceptance developments in Europe and the rest of the World. (author)

  12. Control rod drive

    International Nuclear Information System (INIS)

    Hawke, B.C.

    1986-01-01

    A reactor core, one or more control rods, and a control rod drive are described for selectively inserting and withdrawing the one or more control rods into and from the reactor core, which consists of: a support structure secured beneath the reactor core; control rod positioning means supported by the support structure for movably supporting the control rod for movement between a lower position wherein the control rod is located substantially beneath the reactor core and an upper position wherein at least an upper portion of the control rod extends into the reactor core; transmission means; primary drive means connected with the control rod positioning means by the transmission means for positioning the control rod under normal operating conditions; emergency drive means for moving the control rod from the lower position to the upper position under emergency conditions, the emergency drive means including a weight movable between an upper and a lower position, means for movably supporting the weight, and means for transmitting gravitational force exerted on the weight to the control rod positioning means to move the control rod upwardly when the weight is pulled downwardly by gravity; the transmission means connecting the control rod positioning means with the emergency drive means so that the primary drive means effects movement of the weight and the control rod in opposite directions under normal conditions, thus providing counterbalancing to reduce the force required for upward movement of the control rod under normal conditions; and restraint means for restraining the fall of the weight under normal operating conditions and disengaging the primary drive means to release the weight under emergency conditions

  13. Managing bundled payments.

    Science.gov (United States)

    Draper, Andrew

    2011-04-01

    Results of Medicare's ACE demonstration project and Geisinger Health System's ProvenCare initiative provide insight into the challenges hospitals will face as bundled payment proliferates. An early analysis of these results suggests that hospitals would benefit from bringing full automation using clinical IT tools to bear in their efforts to meet these challenges. Other important factors contributing to success include board and physician leadership, organizational structure, pricing methodology for bidding, evidence-based medical practice guidelines, supply cost management, process efficiency management, proactive and aggressive case management, business development and marketing strategy, and the financial management system.

  14. Muon bundles from the Universe

    Directory of Open Access Journals (Sweden)

    Kankiewicz P.

    2018-01-01

    Full Text Available Recently the CERN ALICE experiment, in its dedicated cosmic ray run, observed muon bundles of very high multiplicities, thereby confirming similar findings from the LEP era at CERN (in the CosmoLEP project. Significant evidence for anisotropy of arrival directions of the observed high multiplicity muonic bundles is found. Estimated directionality suggests their possible extragalactic provenance. We argue that muonic bundles of highest multiplicity are produced by strangelets, hypothetical stable lumps of strange quark matter infiltrating our Universe.

  15. Infinitesimal bundles and projective relativity

    International Nuclear Information System (INIS)

    Evans, G.T.

    1973-01-01

    An intrinsic and global presentation of five-dimensional relativity theory is developed, in which special coordinate conditions are replaced by conditions of Lie invariance. The notion of an infinitesimal bundle is introduced, and the theory of connexions on principal bundles is extended to infinitesimal bundles. Global aspects of projective relativity are studied: it is shown that projective relativity can describe almost any space-time. In particular, it is not necessary to assume that the electromagnetic field have a global potential. (author)

  16. Fuel rod leak detector

    International Nuclear Information System (INIS)

    Womack, R.E.

    1978-01-01

    A typical embodiment of the invention detects leaking fuel rods by means of a radiation detector that measures the concentration of xenon-133 ( 133 Xe) within each individual rod. A collimated detector that provides signals related to the energy of incident radiation is aligned with one of the ends of a fuel rod. A statistically significant sample of the gamma radiation (γ-rays) that characterize 133 Xe is accumulated through the detector. The data so accumulated indicates the presence of a concentration of 133 Xe appropriate to a sound fuel rod, or a significantly different concentration that reflects a leaking fuel rod

  17. Analysis of Double-encapsulated Fuel Rods

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Medvedev, Pavel G [Idaho National Laboratory; Novascone, Stephen Rhead [Idaho National Laboratory; Perez, Danielle Marie [Idaho National Laboratory; Williamson, Richard L [Idaho National Laboratory

    2014-09-01

    In an LWR fuel rod, the cladding encapsulates the fuel, contains fission products, and transfers heat directly to the water coolant. In some situations, it may be advantageous to separate the cladding from the coolant through use of a secondary cladding or capsule. This may be done to increase confidence that the fuel or fission products will not mix with the coolant, to provide a mechanism for controlling the rod temperature, or to place multiple experimental rodlets within a single housing. With an axisymmetric assumption, it is possible to derive closed-form expressions for the temperature profile in a fuel rod using radially-constant thermal conductivity in the fuel. This is true for both a traditional fuel-cladding rod and a double-encapsulated fuel (fuel, cladding, capsule) configuration. Likewise, it is possible to employ a fuel performance code to analyse both a traditional and a double-encapsulated fuel. In the case of the latter, two sets of gap heat transfer conditions must be imposed. In this work, we review the equations associated with radial heat transfer in a cylindrical system, present analytic and computational results for a postulated power and gas mixture history for IFA-744, and describe the analysis of the AFC-2A, 2B metallic fuel alloy experiments at the Advanced Test Reactor, including the effect of a release of fission products into the cladding-capsule gap. The computational results for these two cases were obtained using BISON, a fuel performance code under development at Idaho National Laboratory.

  18. 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

  19. Recycle of LWR actinides to an IFR

    International Nuclear Information System (INIS)

    Pierce, R.D.; Ackerman, J.P.; Johnson, G.K.; Mulcahey, T.P.; Poa, D.S.

    1991-01-01

    Large quantities of actinide elements are present in irradiated light water reactor fuel that is stored throughout the world. Because of the high fission to capture ratio for the transuranium (TRU) elements with the high energy neutrons in metal-fueled integral fast reactors (IFR), that reactor can consume these elements effectively. The stored fuel may represent valuable resource for the expanding application of fast power reactors. In addition, the removal of TRU elements from spent LWR fuel has the potential for increasing the capacity of high level waste facilities by reducing the heat load and may increase the margin of safety in meeting licensing requirement. Argonne National Laboratory is developing a pyrochemical process, which is compatible with the IFR fuel cycle for the recovery of TRU elements from LWR fuel. The proposed product is a metallic actinide ingot, which can be introduced into the electrorefining step of the IFR process. Two pyrochemical processes, that is, salt transport process and blanket processing study, are discussed in this paper. Also the experimental studies are reported. (K.I.)

  20. Nefness of adjoint bundles for ample vector bundles

    Directory of Open Access Journals (Sweden)

    Hidetoshi Maeda

    1995-11-01

    Full Text Available Let E be an ample vector bundle of rank >1 on a smooth complex projective variety X of dimension n. This paper gives a classification of pairs (X,E whose adjoint bundles K_X+det E are not nef in the case when  r=n-2.