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)
Crossflow between subchannels in a 5 x 5 rod-bundle geometry
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.
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)
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)
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.)
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.
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)
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)
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.
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)
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
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
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
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)
Exploring Bundling Theory with Geometry
Eckalbar, John C.
2006-01-01
The author shows how instructors might successfully introduce students in principles and intermediate microeconomic theory classes to the topic of bundling (i.e., the selling of two or more goods as a package, rather than separately). It is surprising how much students can learn using only the tools of high school geometry. To be specific, one can…
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
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)
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
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.
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
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
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.
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
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.
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.)
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
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)
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
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.)
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
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
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
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
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
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
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
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.
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)
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
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)
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.
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
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
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.
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
Differential geometry bundles, connections, metrics and curvature
Taubes, Clifford Henry
2011-01-01
Bundles, connections, metrics and curvature are the 'lingua franca' of modern differential geometry and theoretical physics. This book will supply a graduate student in mathematics or theoretical physics with the fundamentals of these objects. Many of the tools used in differential topology are introduced and the basic results about differentiable manifolds, smooth maps, differential forms, vector fields, Lie groups, and Grassmanians are all presented here. Other material covered includes the basic theorems about geodesics and Jacobi fields, the classification theorem for flat connections, the
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
Fiber bundle geometry and space-time structure
International Nuclear Information System (INIS)
Nascimento, J.C.
1977-01-01
Within the framework of the geometric formulation of Gauge theories in fiber bundles, the general relation between the bundle connection (Gauge field) and the geometry of the base space is obtained. A possible Gauge theory for gravitation is presented [pt
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)
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
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
Single-Phase Crossflow Mixing in a Vertical Tube Bundle Geometry : An Experimental Study
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
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
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)
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
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)
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.
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
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
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
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)
Laminar simulation of intersubchannel mixing in a triangular nuclear fuel bundle geometry
International Nuclear Information System (INIS)
Zaretsky, A.; Lightstone, M.F.; Tullis, S.
2015-01-01
Highlights: • Quasi-periodic flow was observed through rod-to-wall gaps. • Triangular subchannel flows were fundamentally irregular. • Cross-gap flow was influenced both by local and adjacent cross-gap intensity. • Phase-linking between gaps induced cross-plane peripheral circulation through rod–wall gaps. • Cross-gap flow structure was dependent on subchannel geometry. - Abstract: Predicting temperature distributions in fuel rod bundles is an important component of nuclear reactor safety analysis. Intersubchannel mixing acts to homogenize coolant temperatures thus reducing the likelihood of localized regions of high fuel temperature. Previous research has shown that intersubchannel mixing in nuclear fuel rod bundles is enhanced by a large-scale quasi-periodic energetic fluid motion, which transports fluid on the cross-plane between the narrow gaps connecting subchannels. This phenomenon has also been observed in laminar flows. Unsteady laminar flow simulations were performed in a simplified bundle of three rods with a pipe. Three similar geometries of varying gap width were examined, and a thermal trace was implemented on the first geometry. Thermal mixing was driven by the advection of energy between subchannels by the cross-plane flow. Flow through the rod-to-wall gaps in the wall subchannels alternated with a dominant frequency, particularly when rod-to-wall gaps were smaller than rod-to-rod gaps. Significant phase-linking between rod-to-wall gaps was also observed such that a peripheral circulation occurred through each gap simultaneously. Cross-plane flow through the rod-to-rod gaps in the triangular subchannel was irregular in each case. This was due to the fundamental irregularity of the triangular subchannel geometry. Vortices were continually broken up by cross-plane flow from other gaps due to the odd number of fluid pathways within the central subchannel. Cross-plane flow in subchannel geometries is highly interconnected between gaps. The
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)
Crossed Module Bundle Gerbes; Classification, String Group and Differential Geometry
Jurco, Branislav
2005-01-01
We discuss nonabelian bundle gerbes and their differential geometry using simplicial methods. Associated to any crossed module there is a simplicial group NC, the nerve of the 1-category defined by the crossed module and its geometric realization |NC|. Equivalence classes of principal bundles with structure group |NC| are shown to be one-to-one with stable equivalence classes of what we call crossed module gerbes bundle gerbes. We can also associate to a crossed module a 2-category C'. Then t...
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.
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)
Geometry of Quantum Principal Bundles. Pt. 1
International Nuclear Information System (INIS)
Durdevic, M.
1996-01-01
A theory of principal bundles possessing quantum structure groups and classical base manifolds is presented. Structural analysis of such quantum principal bundles is performed. A differential calculus is constructed, combining differential forms on the base manifold with an appropriate differential calculus on the structure quantum group. Relations between the calculus on the group and the calculus on the bundle are investigated. A concept of (pseudo)tensoriality is formulated. The formalism of connections is developed. In particular, operators of horizontal projection, covariant derivative and curvature are constructed and analyzed. Generalizations of the first Structure Equation and of the Bianchi identity are found. Illustrative examples are presented. (orig.)
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
The differential geometry of higher order jets and tangent bundles
International Nuclear Information System (INIS)
De Leon, M.; Rodrigues, P.R.
1985-01-01
This chapter is devoted to the study of basic geometrical notions required for the development of the main object of the text. Some facts about Jet theory are reviewed. A particular case of Jet manifolds is considered: the tangent bundle of higher order. It is shown that this jet bundle possesses in a canonical way a certain kind of geometric structure, the so called almost tangent structure of higher order, and which is a generalization of the almost tangent geometry of the tangent bundle. Another important fact examined is the extension of the notion of 'spray' to higher order tangent bundles. (Auth.)
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.
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.
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)
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
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)
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)
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
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)
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.
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
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
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.
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
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
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
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
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.)
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
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
International Nuclear Information System (INIS)
Moeller, R.; Tschoeke, H.; Kolodziej, M.
1980-12-01
The operating reliability of nuclear reactors calls for a reliable strength analysis of the highly loaded core elements, one of its prerequisites being the reliable determination of the three-dimensional velocity and temperature fields. To verify thermohydraulics computer programs, extensive local temperature measurements in the rod claddings of the critical bundle zone were performed on a heated 19-rod bundle model with sodium flow and provided with spacer grids (P/D = 1.30; W/D = 1.19). These are the essential results obtained: Outside the spacer grids the azimuthal temperature variations of the side and corner rods are greater by approximately the factor 10 in the bundle geometry under consideration as compared to rods in the central bundle zone. The spacer grids investigated give rise to great local temperature peaks and correspondingly great temperature gradients in the axial and azimuthal directions immediately around the support points. Continuous reduction of a subchannel by rod bowing results in substantial rises of temperature which, however, are limited to the adjacent cladding tube zones. (orig.) [de
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)
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
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
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
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)
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)
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
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
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)
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.
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%.
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
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)
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)
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
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)
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
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)
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
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
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
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)
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
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)
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)
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
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.
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.
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.)
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.
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.
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)
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)
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.)
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
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)
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.
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
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
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.
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.
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)
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
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)
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)
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
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)
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.)
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)
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.
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.
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
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
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.)
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
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
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)
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.
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
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.)
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)
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
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
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.
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)
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.
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)
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
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
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.
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
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
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
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
Study of fuel bundle geometry on inter subchannel flow in a 19 pin wire wrapped bundle
International Nuclear Information System (INIS)
Naveen Raj, M.; Velusamy, D.K.
2015-01-01
In typical sodium cooled fast reactor (SFR) fuel pin bundle, gap between the pins is maintained by helically wound wire wrap around each pin. The presence of wire induces large inter-subchannel transverse flow, eventually promoting mixing and heat transfer. The magnitude of the transverse flow is highly dependent on the various pin-bundle dimensions. Appropriate modeling of these transverse flows in subchannel codes is necessary to predict realistic temperature distribution in pin bundle. Hence, detailed parametric study of transverse flow on pin-bundle geometric parameters has been conducted. The parameters taken for the present study are pin diameter, wire diameter, helical wire pitch and edge gap. Towards this 3-D computational fluid dynamic analysis on a structured mesh of 19 pin bundle is carried out using k-epsilon turbulence model. Periodic oscillations along the primacy flow direction were found in subchannel transverse flow and peripheral pin clad temperatures with periodicity over one pitch length. Based on parametric studies, correlations for transverse flow in central subchannels are proposed. (author)
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
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
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
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
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
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
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
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.
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
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)
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
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.
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
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
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
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)
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)
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)
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
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
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
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.)
Salmingo, Remel Alingalan; Tadano, Shigeru; Abe, Yuichiro; Ito, Manabu
2016-05-12
Treatment for severe scoliosis is usually attained when the scoliotic spine is deformed and fixed by implant rods. Investigation of the intraoperative changes of implant rod shape in three-dimensions is necessary to understand the biomechanics of scoliosis correction, establish consensus of the treatment, and achieve the optimal outcome. The objective of this study was to measure the intraoperative three-dimensional geometry and deformation of implant rod during scoliosis corrective surgery.A pair of images was obtained intraoperatively by the dual camera system before rotation and after rotation of rods during scoliosis surgery. The three-dimensional implant rod geometry before implantation was measured directly by the surgeon and after surgery using a CT scanner. The images of rods were reconstructed in three-dimensions using quintic polynomial functions. The implant rod deformation was evaluated using the angle between the two three-dimensional tangent vectors measured at the ends of the implant rod.The implant rods at the concave side were significantly deformed during surgery. The highest rod deformation was found after the rotation of rods. The implant curvature regained after the surgical treatment.Careful intraoperative rod maneuver is important to achieve a safe clinical outcome because the intraoperative forces could be higher than the postoperative forces. Continuous scoliosis correction was observed as indicated by the regain of the implant rod curvature after surgery.
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
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
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
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)
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
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
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
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
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
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
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)
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
Control rod interaction models for use in 2D and 3D reactor geometries
International Nuclear Information System (INIS)
Bannerman, R.C.
1985-11-01
Control rod interaction models are developed for use in two-dimensional and three-dimensional reactor geometries. These models allow the total worth of any combination of control rods inserted to be determined from the individual worths in conjunction with an algorithm representing interaction effects between them. The validity of the assumptions is demonstrated for fast and thermal systems showing modelling errors of 1#percent# or less in inserted control rod worths. The models are ideally suited for most reactor systems. (UK)
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
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.
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
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
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
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)
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)
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)
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
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.
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
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
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
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
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
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
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
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
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
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)
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)
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
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)
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)
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.
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
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
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.)
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
The effect of microstructure and geometry on the fatigue behaviour of bundle assembly welds
International Nuclear Information System (INIS)
Surette, B.A.; Gabbani, M.
1997-01-01
Cracking of end plates, in the Darlington NGS, was attributed to high-cycle fatigue resulting from flow-induced vibrations. Because the cracks were predominantly associated with the bundle assembly welds and with certain element positions, a program was initiated to study whether the microstructure and geometry of the weld zone affected the fatigue behaviour of the assembly welds. Assembly weld samples were subjected to different heat treatments, resulting in different microstructures of the weld zone. Results of fatigue testing suggest that heat treatment of the welds (i.e., microstructure) had little effect on the fatigue life. Assembly welds were also produced with different weld notch geometries, and compared with samples having notches produced by machining (instead of welding). The results of these tests showed that geometry of the weld had a significant effect on fatigue life. However, the geometry of the weld notch required to significantly improve fatigue life is not achievable using the current assembly welding process. A small improvement in fatigue life of welded samples appears possible by increasing the weld diameter. (author)
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)
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)
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
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
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
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
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
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
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
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
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
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)
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
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
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)
International Nuclear Information System (INIS)
Bayoumi, M.A.A.
1976-10-01
A bibliographic study shows that the experimental studies examined, have been developed to understand the phenomenon acting on the mixing between the sub-channels of which geometries are such these of rod bundles used in some nuclear reactors. Experimental devices and tests have been developed to study the influence of the following parameters, operating conditions, pressure, flow rate, power brought to the bundle and inlet temperature on the distribution of flow rates and vapor content among the different sub-channels. By means of non isokinetic sampling, one has determined the enthalpy of the fluid participating to the mixing between the communicating sub-channels and it has been shown that the value of this enthalpy depends strongly on the type of fluid flow and that this enthalpy cannot be either the enthalpy of one of the two sub-channels, nor (always) an average of these two enthalpies. The experimental results have been compared with calculations developed with the code FLICA, concerning the mass velocity distribution, the exchange term of linear momentum, and the variation of the transversal enthalpy with regard to the type of fluid flow. A study of local void ratio measurement, by means of optical probes, has been proposed. The present study has been carried out with a smooth geometry [fr
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
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
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
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
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
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
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)
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.
DEFF Research Database (Denmark)
Sommer, Stefan Horst; Svane, Anne Marie
2017-01-01
distributions. We discuss a factorization of the frame bundle projection map through this bundle, the natural sub-Riemannian structure of the frame bundle, the effect of holonomy, and the existence of subbundles where the Hormander condition is satisfied such that the Brownian motions have smooth transition......We discuss the geometric foundation behind the use of stochastic processes in the frame bundle of a smooth manifold to build stochastic models with applications in statistical analysis of non-linear data. The transition densities for the projection to the manifold of Brownian motions developed...... in the frame bundle lead to a family of probability distributions on the manifold. We explain how data mean and covariance can be interpreted as points in the frame bundle or, more precisely, in the bundle of symmetric positive definite 2-tensors analogously to the parameters describing Euclidean normal...
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
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.
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)
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)
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)
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
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
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
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)
International Nuclear Information System (INIS)
Falzetti, L.; Meneghello, S.; Pezzilli, M.
1979-01-01
Two sets of experiments have been performed in sodium with two seven pin electrically heated bundles: the first with a nominal arrangement, the second with one dummy pin enlarged 20% in diameter in peripheral position. In this paper a rapid review of experimental results and theoretical works, related to the temperature distribution in these geometries, is presented together with a short description of the developed test section technology
Salmingo, Remel A; Tadano, Shigeru; Fujisaki, Kazuhiro; Abe, Yuichiro; Ito, Manabu
2012-05-01
Scoliosis is defined as a spinal pathology characterized as a three-dimensional deformity of the spine combined with vertebral rotation. Treatment for severe scoliosis is achieved when the scoliotic spine is surgically corrected and fixed using implanted rods and screws. Several studies performed biomechanical modeling and corrective forces measurements of scoliosis correction. These studies were able to predict the clinical outcome and measured the corrective forces acting on screws, however, they were not able to measure the intraoperative three-dimensional geometry of the spinal rod. In effect, the results of biomechanical modeling might not be so realistic and the corrective forces during the surgical correction procedure were intra-operatively difficult to measure. Projective geometry has been shown to be successful in the reconstruction of a three-dimensional structure using a series of images obtained from different views. In this study, we propose a new method to measure the three-dimensional geometry of an implant rod using two cameras. The reconstruction method requires only a few parameters, the included angle θ between the two cameras, the actual length of the rod in mm, and the location of points for curve fitting. The implant rod utilized in spine surgery was used to evaluate the accuracy of the current method. The three-dimensional geometry of the rod was measured from the image obtained by a scanner and compared to the proposed method using two cameras. The mean error in the reconstruction measurements ranged from 0.32 to 0.45 mm. The method presented here demonstrated the possibility of intra-operatively measuring the three-dimensional geometry of spinal rod. The proposed method could be used in surgical procedures to better understand the biomechanics of scoliosis correction through real-time measurement of three-dimensional implant rod geometry in vivo.
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
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
International Nuclear Information System (INIS)
Bottoni, M.; Lyczkowski, R.; Ahuja, S.
1995-01-01
Numerical simulation of subcooled boiling in one-dimensional geometry with the Homogeneous Equilibrium Model (HEM) may yield difficulties related to the very low sonic velocity associated with the HEM. These difficulties do not arise with subcritical flow. Possible solutions of the problem include introducing a relaxation of the vapor production rate. Three-dimensional simulations of subcooled boiling in bundle geometry typical of fast reactors can be performed by using two systems of conservation equations, one for the HEM and the other for a Separated Phases Model (SPM), with a smooth transition between the two models
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)
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
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)
On the geometry of the fuel rod supports concerning a fretting wear failure
International Nuclear Information System (INIS)
Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Kang-Hee
2008-01-01
Geometrical conditions of spacer grid springs and dimples of a light water reactor fuel assembly are studied in this paper concerning a fuel rod's fretting wear failure. In this framework, the springs/dimples are categorized from the aspects of their orientation with respect to the fuel axis and the contact types. Possible motions on the contacts between the springs/dimples and fuel rods are estimated by conducting a flow-induced vibration test. Features of the wear scar and depth are investigated by independent fretting wear tests carried out with spring and dimple specimens of typical contact geometries. It is also attempted here to apply the contact mechanics theory to a fuel fretting wear analysis such as the prediction of a wear depth profile and its rate, which is influenced by the contact shape of the springs/dimples. It is shown that the theory can be applied to a dimensional control of a coining for the springs/dimples, which is usually carried out in a thin plate fabrication. From the results, the necessary conditions for a spring/dimple geometry for restraining a fretting wear failure are discussed
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
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
Symplectic geometry on moduli spaces of holomorphic bundles over complex surfaces
Khesin, Boris; Rosly, Alexei
2000-01-01
We give a comparative description of the Poisson structures on the moduli spaces of flat connections on real surfaces and holomorphic Poisson structures on the moduli spaces of holomorphic bundles on complex surfaces. The symplectic leaves of the latter are classified by restrictions of the bundles to certain divisors. This can be regarded as fixing a "complex analogue of the holonomy" of a connection along a "complex analogue of the boundary" in analogy with the real case.
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)
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
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.
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
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
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
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
Heterotic Non-Kähler Geometries via Polystable Bundles on Calabi-Yau Threefolds
DEFF Research Database (Denmark)
Andreas, Bjorn; Garcia Fernandez, Mario
2012-01-01
In arXiv:1008.1018 it is shown that a given stable vector bundle V on a Calabi-Yau threefold X which satisfies c_2(X) = c_2(V ) can be deformed to a solution of the Strominger system and the equations of motion of heterotic string theory. In this note we extend this result to the polystable case...
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
International Nuclear Information System (INIS)
Herkenrath, H.; Hufschmidt, W.; Jung, U.; Weckermann, F.
1981-01-01
The enthalpy- and mass-flow-distribution at the outlet of two different 16-rod cluster test sections with uniform heating in axial and radial direction under steady state conditions has been measured for the first time by simultaneous sampling of 5 from 6 present characteristic subchannels in the bundle using the isokinetic technique and analysing the outlet quantities by a calorimetic method. The test-sections are provided with typical geometrical configurations for BWR s (70 bars; test section PELCO-S) and PWR s (160 bars; test-section EUROP). The latter has also been tested under BWR conditions (70 bars) to study the influence of geometry and pressure. The results showed the abnormal behaviour of the corner subchannel under BWR typical conditions (70 bars) which could not be found for PWR conditions (160 bars) and which is only an effect of pressure and not of geometry. The analysis of the experimental data confirms the usefullness of the subchannel sampling technique for the better understanding of the complex thermohydraulic phenomena under two-phase flow conditions in multirod bundles. Calculations of subchannel resistance coefficients for both types of spacers under one-phase flow conditions have been made with a special sub-structure method which showed a rather high local value of the corner subchannel. With the local drag coefficents the total resistance of the spacer has been evaluated and agreed well with measured values under adiabatic conditions. The measured subchannel data permit a direct valuation and examination of respective computer codes in a fundamental manner which are, however, not subject of this report
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
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).
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.)
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
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
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)
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
International Nuclear Information System (INIS)
Chun, S. Y.; Mun, S. K.; Park, J. K.; Yang, S. K.; Jung, M. K.
1999-01-01
The relationship between the boiling length and the CHF in annulus geometry with uniformly heated rod has been studied. In this study the CHF data under pressure of 0.57∼15.01 MPa, flow rate of 200∼650 kg/m 2 s, inlet subcooling of 85∼353 kJ/kg and exit quality of 0.106∼0.536 have been applied. As a result of examining the flow pattern over the heated section, all of the CHF data were the dryout type CHF in annular flow and the locations of the churn to annular flow transition moved down stream of the heated section with increasing the pressure. The effect of pressure on the boiling length under the CHF conditions showed the trends similar to the effect of pressure on the CHF. The relationship between the non-dimensional CHF, q CHF and mass flux taking into account of the boiling length, G ( L h / L B ) indicated the linear relationship without scatter and regardless of pressure and inlet subcooling. The CHF calculated by using the relationship between the non-dimensionless CHF, q CHF and mass flux, G ( L h / L B ) predicted very well the experimental CHF data with the pressure dependence
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.
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
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.
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.
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
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.
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
Comparison of rod-ejection transient calculations in hexagonal-Z geometry
International Nuclear Information System (INIS)
Knight, M.P.; Brohan, P.; Finnemann, H.; Huesken, J.
1995-01-01
This paper proposes a set of 3-dimensional benchmark rod ejection problems for a VVER reactor, based on the well-known NEACRP PWR rod-ejection problems defined by Siemens/KWU. Predictions for these benchmarks derived using three hexagonal-z nodal transient codes, the PANTHER code of Nuclear Electric, the HEXTIME code of Siemens/KWU, and the DYN3D code of FZ-Rossendorf are presented and compared
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
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.
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
International Nuclear Information System (INIS)
Kiteley, J.C.; Carver, M.B.; Zhou, Q.N.
1993-01-01
The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting critical heat flux (CHF) at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is the only tool available to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries. 28 refs., 12 figs
International Nuclear Information System (INIS)
Carver, M.B.; Kiteley, J.C.; Zhou, R.Q.N.; Junop, S.V.; Rowe, D.S.
1995-01-01
The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of Canada uranium deuterium (CANDU) pressurized heavy water reactor fuel channels and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting CHF at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental database. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. The numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology are discussed. The evolutionary validation plan is also discussed and early validation exercises are summarized. More recent validation exercises in standard and nonstandard geometries are emphasized
International Nuclear Information System (INIS)
Carver, M.B.; Kiteley, J.C.; Zhou, R.Q.N.; Junop, S.V.; Rowe, D.S.
1993-01-01
The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of prediting CHF at these local conditions, makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries
International Nuclear Information System (INIS)
Vasiliev, A.D.; Kobelev, G.V.; Astafieva, V.O.
2007-01-01
Radiative heat transfer is very important in different fields of mechanical engineering and related technologies including nuclear reactors, heat transfer in furnaces, aerospace, different high-temperature assemblies. In particular, in the course of a hypothetical severe accident at PWR-type nuclear reactor the temperatures inside the reactor vessel reach high values at which taking into account of radiative heat exchange between the structures of reactor (including core and other reactor vessel elements) gets important. Radiative heat transfer dominates the late phase of severe accident because radiative heat fluxes (proportional to T4, where T is the temperature) are generally considerably higher than convective and conductive heat fluxes in a system. In particular, heat transfer due to radiation determines the heating and degradation of the core and surrounding steel in-vessel structures and finally influences the composition, temperature and mass of materials pouring out of the reactor vessel after its loss of integrity. Existing models of radiative heat exchange use many limitations and approximations: approximate estimation of view factors and beam lengths; the geometry change in the course of the accident is neglected; the database for emissivities of materials is not complete; absorption/emission by steam-noncondensable medium is taken into account approximately. The module MRAD was developed in this paper to model the radiative heat exchange in rod-like geometry typical of PWR-type reactor. Radiative heat exchange is computed using dividing on zones (zonal method) as in existing radiation models implemented to severe accident numerical codes such as ICARE, SCDAP/RELAP, MELCOR but improved in following aspects: new approach to evaluation of view factors and mean beam length; detailed evaluation of gas absorptivity and emissivity; account of effective radiative thermal conductivity for the large core; account of geometry modification in the course of severe
International Nuclear Information System (INIS)
Fulfs, H.; Katsaounis, A.; Kreubig, M.; Minden, C. von; Orlowski, R.
1980-01-01
The results will be described in exemplary presentations completely and concluding. The experimental examination of the steady state simularity of critical heat flux (CHF) in freon 12 and water at identical PWR-5 x 15-rod bundles will show that hot rod/hot channels position as well as CHF can be transformed from model to original fluid with good accuracy. The investigated mass flow and power transients (only in freon 12) point out a definite influence of initial and boundary conditions on CHF and CHF time delay at changing rates higher than 10 to 20%/s. On the contrary simulation of primary pump failure (LOFA) shows no or only small improvement in CHF behaviour while a coupled Scram prevents from reaching the boiling crisis. (orig.) [de
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.)
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)
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
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
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
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.
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)
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)
International Nuclear Information System (INIS)
Niceno, B.; Sato, Y.; Badillo, A.; Andreani, M.
2010-01-01
In this paper we describe current activities on the project Multi-Scale Modeling and Analysis of convective boiling (MSMA), conducted jointly by the Paul Scherrer Institute (PSI) and the Swiss Nuclear Utilities (Swissnuclear). The long-term aim of the MSMA project is to formulate improved closure laws for Computational Fluid Dynamics (CFD) simulations for prediction of convective boiling and eventually of the Critical Heat Flux (CHF). As boiling is controlled by the competition of numerous phenomena at various length and time scales, a multi-scale approach is employed to tackle the problem at different scales. In the MSMA project, the scales on which we focus range from the CFD scale (macro-scale), bubble size scale (meso-scale), liquid micro-layer and triple interline scale (micro-scale), and molecular scale (nano-scale). The current focus of the project is on micro- and meso- scales modeling. The numerical framework comprises a highly efficient, parallel DNS solver, the PSI-BOIL code. The code has incorporated an Immersed Boundary Method (IBM) to tackle complex geometries. For simulation of meso-scales (bubbles), we use the Constrained Interpolation Profile method: Conservative Semi-Lagrangian 2nd order (CIP-CSL2). The phase change is described either by applying conventional jump conditions at the interface, or by using the Phase Field (PF) approach. In this work, we present selected results for flows in complex geometry using the IBM, selected bubbly flow simulations using the CIP-CSL2 method and results for phase change using the PF approach. In the subsequent stage of the project, the importance of effects of nano-scale processes on the global boiling heat transfer will be evaluated. To validate the models, more experimental information will be needed in the future, so it is expected that the MSMA project will become the seed for a long-term, combined theoretical and experimental program
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)
Diversion cross-flow mixing at the inlet of a simulated rod bundle using a gamma camera
International Nuclear Information System (INIS)
Sedaghat, A.; Macduff, R.; Castellana, F.
1986-01-01
The prediction of diversion cross-flow and turbulent mixing interests reactor vendors and nuclear fuel suppliers because of the effect on critical heat flux. In single-phase flow with uniform inlet conditions, flow diversion occurs primarily near the inlet. Prior work by Bowring and Levy and Lahey estimated diversion length by comparing the axial pressure differential at the channel exit using isokinetic (natural flow split) and nonisokinetic (forced flow split) sampling and by using a mathematical model. The present work, sponsored by Exxon Nuclear Company, Inc., represents the first study in which flow distribution and diversion cross flow were investigated at the inlet of a clean geometry. The parameters investigated were diversion length and the effective cross-flow velocity was determined by analysis. The results of this work were compared to theoretical values predicted by the COBRA IIIC subchannel computer code. The difference between experimental data and COBRA IIIC suggests that a more comprehensive transverse momentum balance is desired as mass flux ratios become large. The inclusion of transverse inertia and acceleration terms in the transverse momentum balance become important
Sossinsky, A B
2012-01-01
The book is an innovative modern exposition of geometry, or rather, of geometries; it is the first textbook in which Felix Klein's Erlangen Program (the action of transformation groups) is systematically used as the basis for defining various geometries. The course of study presented is dedicated to the proposition that all geometries are created equal--although some, of course, remain more equal than others. The author concentrates on several of the more distinguished and beautiful ones, which include what he terms "toy geometries", the geometries of Platonic bodies, discrete geometries, and classical continuous geometries. The text is based on first-year semester course lectures delivered at the Independent University of Moscow in 2003 and 2006. It is by no means a formal algebraic or analytic treatment of geometric topics, but rather, a highly visual exposition containing upwards of 200 illustrations. The reader is expected to possess a familiarity with elementary Euclidean geometry, albeit those lacking t...
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
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
Indian Academy of Sciences (India)
. In the previous article we looked at the origins of synthetic and analytic geometry. More practical minded people, the builders and navigators, were studying two other aspects of geometry- trigonometry and integral calculus. These are actually ...
Prasolov, V V
2015-01-01
This book provides a systematic introduction to various geometries, including Euclidean, affine, projective, spherical, and hyperbolic geometries. Also included is a chapter on infinite-dimensional generalizations of Euclidean and affine geometries. A uniform approach to different geometries, based on Klein's Erlangen Program is suggested, and similarities of various phenomena in all geometries are traced. An important notion of duality of geometric objects is highlighted throughout the book. The authors also include a detailed presentation of the theory of conics and quadrics, including the theory of conics for non-Euclidean geometries. The book contains many beautiful geometric facts and has plenty of problems, most of them with solutions, which nicely supplement the main text. With more than 150 figures illustrating the arguments, the book can be recommended as a textbook for undergraduate and graduate-level courses in geometry.
International Nuclear Information System (INIS)
Ninokata, Hisashi; Sadatomi, Michio; Okawa, Tomio
2003-01-01
In order to establish a key technology to realize advanced BWR fuel designs, a three-year project of the advanced subchannel analysis code development had been started since 2002. The five dominant factors involved in the boiling transitional process in the fuel bundles were focused. They are, (1) inter-subchannel exchanges, (2) influences of obstacles (3) dryout of liquid film, (4) transition of two-phase flow regimes and (5) deposition of droplets. It has been recognized that present physical models or constitutive equations in subchannel formulations need to be improved so that they include geometrical effects in the fuel bundle design more mechanistically and universally. Through reviewing literatures and existent experimental results, underlying elementary processes and geometrical factors that are indispensable for improving subchannel codes were identified. The basic strategy that combines numerical and experimental approaches was proposed aiming at establishment of mechanistic models for the five dominant factors. In this paper, the present status of methodologies for detailed two-phase flow studies has been summarized. According to spatial scales of focused elementary processes, proper numerical approaches were selected. For some promising numerical approaches, preliminary calcitonins were performed for assessing their applicability to investigation of elementary processes involved in the boiling transition. (author)
International Nuclear Information System (INIS)
Bottoni, M.; Dorr, B.; Homann, C.; Struwe, D.
1983-07-01
The computer programme BACCHUS implemented at KfK includes a steady-state version, a two-dimensional and a three-dimensional transient single-phase flow version describing the thermal-hydraulic behaviour of the coolant (sodium or water) in bundle geometry under nominal or accident conditions. All versions are coupled with a pin model describing the temperature distribution in fuel (or electrical heaters) and cladding. The report describes the programme from the viewpoints of the geometrical model, the mathematical foundations and the numerical treatment of the basic equations. Although emphasis is put on the three-dimensional version, the two-dimensional and the steady state versions are also documented in self-consistent sections. (orig.) [de
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
Pedoe, Dan
1988-01-01
""A lucid and masterly survey."" - Mathematics Gazette Professor Pedoe is widely known as a fine teacher and a fine geometer. His abilities in both areas are clearly evident in this self-contained, well-written, and lucid introduction to the scope and methods of elementary geometry. It covers the geometry usually included in undergraduate courses in mathematics, except for the theory of convex sets. Based on a course given by the author for several years at the University of Minnesota, the main purpose of the book is to increase geometrical, and therefore mathematical, understanding and to he
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
CFD analysis of flow and heat transfer in Canadian supercritical water reactor bundle
International Nuclear Information System (INIS)
Podila, K.; Rao, Y.F.
2015-01-01
Highlights: • Flow and heat transfer in SCWR fuel bundle design by AECL is studied using CFD. • Bare-rod bundle geometry is tested at 23.5, 25 and 28 MPa using STAR-CCM+ code. • SST k–ω low-Re model was used to study occurrence of heat transfer deterioration. - Abstract: Within the Gen-IV International Forum, AECL is leading the effort in developing a conceptual design for the Canadian SCWR. AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for occurrence of HTD in the supercritical bundle flows. In the current investigation, bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The low-Reynolds number modification of SST k–ω turbulence model along with y + < 1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5 MPa)
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
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
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
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.)
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.)
Indian Academy of Sciences (India)
algebraic geometry but also in related fields like number theory. ... every vector bundle on the affine space is trivial. (equivalently ... les on a compact Riemann surface to unitary rep- ... tial geometry and topology and was generalised in.
Effect of 17 x 17 fuel assembly geometry on interchannel thermal mixing
International Nuclear Information System (INIS)
Motley, F.E.; Wenzell, A.H.; Cadek, F.F.
1975-01-01
A test to determine the value of the thermal diffusion coefficient (TDC) in the 17 x 17 fuel assembly geometry was conducted. The test section was a 5 x 5 rod bundle with a radial power difference of 4.5 to 1. The rod OD and pitch are identical to the 17 x 17 fuel assembly, as is the mixing vane grid design. The value of thermal diffusion coefficient (TDC) was determined by matching the experimental exit enthalpy distribution to that predicted by the THINC computer code. The mean value of TDC for the 17 x 17 fuel assembly geometry is TDC = .059. 6 references
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)
Visual observations of a degraded bundle of irradiated fuel: the Phebus FPT1 test
International Nuclear Information System (INIS)
Barrachin, M.; Bottomley, P.D.
1999-01-01
The international Phebus-FP (Fission Product) project is managed by the Institut de Protection et Surete Nucleaire in collaboration with Electricite de France (EDF), the European Commission (EC), the USNRC (USA), COG (Canada), NUPEC and JAERI (Japan), KAERI (South Korea), PSI and HSK (Switzerland). It is designed to measure the source-term and to study the degradation of irradiated UO 2 fuel in conditions typical of a severe loss of coolant accident in a pressurised water reactor (PWR). In the first test (FPT0), performed in December '93, a bundle of 20 fresh fuel rods and a central Ag-In-Cd control rod underwent a short 15-day irradiation to generate fission products before testing in the Phebus reactor in Cadarache. The second test (FPT1) was performed in July '96, in the same conditions and geometry, but using irradiated fuel (-23 GWd/tU). In the FPT1 test, the bundle was heated to an estimated 3000 K over a period of 30 minutes in order to induce a substantial liquefaction of the bundle. After the test, the bundle was embedded in epoxy and cut at different levels to investigate the mechanisms of the core degradation. This paper reports the visual observations of the degraded FPT1 bundle, very preliminary interpretations about the scenario of degradation and a comparison between the behaviour of the fuel in the FPT0 and FPT1 tests. (author)
Results of the QUENCH-12 experiment on reflood of a VVER-type bundle
International Nuclear Information System (INIS)
Stuckert, J.; Grosse, M.; Heck, M.; Schanz, G.; Sepold, L.; Stegmaier, U.; Steinbrueck, M.
2008-09-01
The QUENCH experiments are to investigate the hydrogen source term resulting from the water injection into an uncovered core of a Light-Water Reactor. The QUENCH test bundle with a total length of approximately 2.5 m usually consists of 21 fuel rod simulators of Western PWR (Pressurized Water Reactor) geometry. The QUENCH-12 test bundle, however, which was set up to investigate the effects of VVER materials and bundle geometry (hexagonal lattice) on core reflood consisted of 31 fuel rod simulators. 18 rods of which were electrically heated using tungsten heaters in the rod center. All claddings, corner rods and grid spacers were made of Zr1%Nb (E110) and the shroud of Zr2.5%Nb (E125). For comparison, the QUENCH-06 test (ISP-45) with Western PWR geometry (square lattice) was chosen as reference. QUENCH-12 conducted at the Forschungszentrum Karlsruhe (FZK, Karlsruhe Research Center) on 27 September, 2006 in the frame of the EC-supported ISTC program 1648.2 was proposed by FZK together with RIAR Dimitrovgrad and IBRAE Moscow (Russia), and supported by pretest calculations performed by PSI (Switzerland) and the Kurchatov Institute Moscow (Russia) together with IRSN Cadarache (France). It had been preceded by a low-temperature (maximum 1073 K) pretest on 25 August, 2006 to characterize the bundle thermal hydraulic performance and to provide data to assess the code models used for pretest calculational support. After a stabilization period at 873 K pre-oxidation took place at ∝1470 K for ∝3400 s to achieve a maximum oxide thickness of about 200 μm. A transient phase followed with a temperature rise to ∝2050 K. Then quenching of the bundle by a water flow of 48 g/s was initiated cooling the bundle to ambient temperature in ∝5 min. Following reflood initiation, a moderate temperature excursion of ∝50 K was observed, over a longer period than in QUENCH-06. The temperatures at elevations between 850 mm and 1050 mm exceeded the melting temperature of β-Zr, i
Results of the QUENCH-12 experiment on reflood of a VVER-type bundle
Energy Technology Data Exchange (ETDEWEB)
Stuckert, J.; Grosse, M.; Heck, M.; Schanz, G.; Sepold, L.; Stegmaier, U.; Steinbrueck, M. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Materialforschung, Programm Nukleare Sicherheitsforschung; Goryachev, A.; Ivanova, I. [RIAR (FSUE SSC-RIAR) Dimitrovgrad (Russian Federation)
2008-09-15
The QUENCH experiments are to investigate the hydrogen source term resulting from the water injection into an uncovered core of a Light-Water Reactor. The QUENCH test bundle with a total length of approximately 2.5 m usually consists of 21 fuel rod simulators of Western PWR (Pressurized Water Reactor) geometry. The QUENCH-12 test bundle, however, which was set up to investigate the effects of VVER materials and bundle geometry (hexagonal lattice) on core reflood consisted of 31 fuel rod simulators. 18 rods of which were electrically heated using tungsten heaters in the rod center. All claddings, corner rods and grid spacers were made of Zr1%Nb (E110) and the shroud of Zr2.5%Nb (E125). For comparison, the QUENCH-06 test (ISP-45) with Western PWR geometry (square lattice) was chosen as reference. QUENCH-12 conducted at the Forschungszentrum Karlsruhe (FZK, Karlsruhe Research Center) on 27 September, 2006 in the frame of the EC-supported ISTC program 1648.2 was proposed by FZK together with RIAR Dimitrovgrad and IBRAE Moscow (Russia), and supported by pretest calculations performed by PSI (Switzerland) and the Kurchatov Institute Moscow (Russia) together with IRSN Cadarache (France). It had been preceded by a low-temperature (maximum 1073 K) pretest on 25 August, 2006 to characterize the bundle thermal hydraulic performance and to provide data to assess the code models used for pretest calculational support. After a stabilization period at 873 K pre-oxidation took place at {proportional_to}1470 K for {proportional_to}3400 s to achieve a maximum oxide thickness of about 200 {mu}m. A transient phase followed with a temperature rise to {proportional_to}2050 K. Then quenching of the bundle by a water flow of 48 g/s was initiated cooling the bundle to ambient temperature in {proportional_to}5 min. Following reflood initiation, a moderate temperature excursion of {proportional_to}50 K was observed, over a longer period than in QUENCH-06. The temperatures at elevations
International Nuclear Information System (INIS)
Anklam, T.M.; White, M.D.
1981-01-01
Experimental data is reported from a series of quasi-steady-state two-phase mixture level swell and void fraction distribution tests. Testing was performed at ORNL in the Thermal Hydraulic Test Facility - a large electrically heated test loop configured to produce conditions similar to those expected in a small break loss of coolant accident. Pressure was varied from 2.7 to 8.2 MPa and linear power ranged from 0.33 to 1.95 kW/m. Mixture swell was observed to vary linearly with the total volumetric vapor generation rate over the power range of primary interest in small break analysis. Void fraction data was fit by a drift-flux model and both the drift-velocity and concentration parameter were observed to decrease with increasing pressure
Energy Technology Data Exchange (ETDEWEB)
Stender, Anthony [Iowa State Univ., Ames, IA (United States)
2013-01-01
Gold nanoparticles, particularly those with an anisotropic shape, have become a popular optical probe for experiments involving work on the nanoscale. However, to carry out such delicate and intricate experiments, it is first necessary to understand the detailed behavior of individual nanoparticles. In this series of experiments, optical and electron microscopy were utilized for the characterization of individual nanoparticles and small assemblies of nanoparticles. In the first experiment, gold nanorods were investigated. Single, isolated nanorods exhibit two maxima of localized surface plasmon resonance (LSPR), which are associated with the two nanorod axes. Upon the physical rotation of a nanorod at one of its LSPR wavelengths under polarized illumination, the optical behavior varies in a sinusoidal fashion. A dimer of nanorods exhibits optical behavior quite similar to a nanorod, except the LSPR maxima are shifted and broader. Under differential interference contrast (DIC) microscopy, a pair of nanorods separated by a distance below the diffraction limit can be distinguished from a single nanorod due to its optical behavior upon rotation. Dark field microscopy is unable to distinguish the two geometries. For the second set of experiments, the optical behavior of single gold nanorods at non-plasmonic wavelengths was investigated. The same nanorod was rotated with respect to a polarized light source under DIC, dark field, and polarized light microscopy. DIC microscopy was found to produce diffraction pattern peaks at non-plasmonic wavelengths, which could be altered by adjusting the setting of the polarizer. In the third set of experiments, the optical behavior of a single gold dumbbell and several simple dumbbell geometries were investigated with microscopy and simulations. The single dumbbell displayed behavior quite similar to that of a nanorod, but dumbbells exhibit a shift in both LSPR wavebands. Moreover, the shape of dumbbell particles allows them to
International Nuclear Information System (INIS)
Kobzar, L.L.; Oleksyuk, D.A.
2001-01-01
The results of experiments executed in RRC 'Kurchatov Institute on the thermal-physical critical facility SVD are presented herein. The experiments modeled the drawing of two fuel rods to each other till touching WWER-1000 reactor in FA. The experimental model is a 7-rod bundle with the heated length of 1 m. The primary goal of experiments was to acquire the quantitative factors of the reduction in the critical heat fluxes as contrasted to the basic model (without disturbances of FA geometry) at the expense of local disturbance of a rod bundle geometry. As it follows from the experiment, the effect of decrease of the critical heat rate depends on combination of regime parameters and it makes 15% in the most unfavorable case (Authors)
Constructing co-Higgs bundles on CP^2
Rayan, Steven
2013-01-01
On a complex manifold, a co-Higgs bundle is a holomorphic vector bundle with an endomorphism twisted by the tangent bundle. The notion of generalized holomorphic bundle in Hitchin's generalized geometry coincides with that of co-Higgs bundle when the generalized complex manifold is ordinary complex. Schwarzenberger's rank-2 vector bundle on the projective plane, constructed from a line bundle on the double cover CP^1 \\times CP^1 \\to CP^2, is naturally a co-Higgs bundle, with the twisted endom...
International Nuclear Information System (INIS)
Zboray, Robert; Kickhofel, John; Damsohn, Manuel; Prasser, Horst-Michael
2011-01-01
Highlights: → Annular flows w/wo functional spacers are investigated by cold neutron imaging. → Liquid film thickness distribution on fuel pins and on spacer vanes is measured. → The influence of the spacers on the liquid film distributions has been quantified. → The cross-sectional averaged liquid hold-up significantly affected by the spacers. → The sapers affect the fraction of the entrained liquid hold up in the gas core. - Abstract: Dryout of the coolant liquid film at the upper part of the fuel pins of a boiling water reactor (BWR) core constitutes the type of heat transfer crisis relevant for the conditions of high void fractions. It is both a safety concern and a limiting factor in the thermal power and thus for the economy of BWRs. We have investigated adiabatic, air-water annular flows in a scaled-up model of two neighboring subchannels as found in BWR fuel assemblies using cold-neutron tomography. The imaging of the double suchannel has been performed at the ICON beamline at the neutron spallation source SINQ at the Paul Scherrer Institute, Switzerland. Cold-neutron tomography is shown here to be an excellent tool for investigating air-water annular flows and the influence of functional spacers of different geometries on such flows. The high-resolution, high-contrast measurements provide the spatial distributions of the coolant liquid film thickness on the fuel pin surfaces as well as on the surfaces of the spacer vanes. The axial variations of the cross-section averaged liquid hold-up and its fraction in the gas core shows the effect of the spacers on the redistribution of the two phases.
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.
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
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.
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)
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)
International Nuclear Information System (INIS)
Catana, Alexandru; Department of Energy Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel
2007-01-01
CANDU 600 nuclear reactors are usually fuelled with STANDARD (STD), 37 rods fuel bundles. Natural uranium (NU) dioxide (UO 2 ), is used as fuel composition. A new fuel bundle geometry called CANFLEX (CFX) with 43 rods is proposed and some new fuel composition are considered. Flexibility is the key word for the attempt to use some different fuel geometries and compositions for CANDU 600 nuclear reactors as well as for innovative ACR-700/1000 nuclear reactors. The fuel bundle considered in this paper is CFX-RU-0.90 that encodes the CANFLEX geometry, recycled dioxide uranium (RU) with 0.90% enrichment. The goal of this proposal is ambitious: a higher average discharge burn-up up to 14000 MWd/tU and, for the same amount of generated electric power, reduction in nuclear fuel fabrication, reduction of spent nuclear fuel radioactive waste and reduction of refueling operational work by using fewer bundles. An improved sub-channel approach for thermal-hydraulic analysis is used in this paper to compute some flow parameters, mainly the pressure drop along the CANDU 600 fuel channel when STD or CFX-RU-0.90 fuel bundles. Also an intermediate CFX-NU fuel bundle are used, for gradual comparison. For CFX-RU- 0.90 four fuel bundle shift refueling scheme is used instead of eight, that will determine different axial power distributions. At the same time radial power distribution is affected by the geometry and by the fuel composition of fuel bundle type used. Some other thermal-hydraulic flow parameters will be influenced, too. One of the most important parameter is pressure drop (PD) along the fuel channel because of its importance in drag force evaluation. We start with an axial power distribution, which is characteristic for a refueling scheme of eight or four fuel bundles on a shift. Comparative results are presented between STD37, CFX-NU CFX-RU-0.90 fuel bundles in a CANDU nuclear reactor operating conditions. Neutron flux distribution analysis shows that four bundle shift
Energy Technology Data Exchange (ETDEWEB)
Bartzis, J G; Megaritou, A; Belessiotis, V
1987-09-01
THEAP-I is a computer code developed in NRCPS `DEMOCRITUS` with the aim to contribute to the safety analysis of the open pool research reactors. THEAP-I is designed for three dimensional, transient thermal/hydraulic analysis of a thermally interacting channel bundle totally immersed into water or air, such as the reactor core. In the present report the mathematical and physical models and methods of the solution are given as well as the code description and the input data. A sample problem is also included, refering to the Greek Research Reactor analysis, under an hypothetical severe loss of coolant accident.
Principal bundles the classical case
Sontz, Stephen Bruce
2015-01-01
This introductory graduate level text provides a relatively quick path to a special topic in classical differential geometry: principal bundles. While the topic of principal bundles in differential geometry has become classic, even standard, material in the modern graduate mathematics curriculum, the unique approach taken in this text presents the material in a way that is intuitive for both students of mathematics and of physics. The goal of this book is to present important, modern geometric ideas in a form readily accessible to students and researchers in both the physics and mathematics communities, providing each with an understanding and appreciation of the language and ideas of the other.
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
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)
CFD simulation of flow and heat transfer in Canadian SCWR bundles
International Nuclear Information System (INIS)
Podila, K.; Rao, Y.F.
2014-01-01
Within the Generation-IV (Gen-IV) International Forum, Atomic Energy of Canada Limited (AECL) is leading the effort in developing a conceptual design for the Canadian supercritical water-cooled reactor (SCWR). AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for possible occurrence of heat transfer deterioration (HTD) in the supercritical bundle flows. In the current investigation, a bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The SST k-ω turbulence model along with y + <1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5MPa). Ultimately, the goal of this study is to aid the development of a criterion for the onset of HTD in the proposed SCWR bundles, which is planned in the next phase of the project. (author)
Simulation of a two phase boiling flow in Poseidon geometry with Astrid steam-water software
International Nuclear Information System (INIS)
Larrauri, D.
1997-01-01
After different validation test runs in tube an annular geometries, the simulation of a subcooled boiling flow in a rod bundle geometry has been achieved with ASTRID Steam-Water software. The experiment we have simulated is the Poseidon experiment. It is a three heating tube geometry. The thermohydraulic conditions of the simulated flow are closed to the DNB conditions. The simulation results are analysed and compared against the available measurements of liquid and wall temperatures. ASTRID Steam-Water behaviour in such a geometry brings satisfaction. The wall and the liquid temperatures are well predicted in the different parts of the flow. The void fraction reaches 40 % in the vicinity of the heating rods. Besides, the evolution of the different calculated variables shows that a three-dimensional simulation gives capital information for the analyse of the physical phenomena involved in this kind of flow. The good results obtained in Poseidon geometry lead us to think about simulating and analyzing rod bundle flows with ASTRID Steam-Water code. (author)
International Nuclear Information System (INIS)
Herkenrath, H.; Hufschmidt, W.
1979-01-01
Experiments based on the subchannel isokinetic technique have been carried out at the JRC of the European Community at Ispra, using a purposely designed 16-rod test section, simulating in a rather accurate way a typical BWR geometry. The adopted system allows the simultaneous determination of mass flow and enthalpy, at the end of the bundle active length, in four characteristic subchannels of the 16-rod lattice. The results show some pronounced flow and enthalpy variations within the bundle, not accurately taken into account by current subchannel codes, such as COBRA-3C. In particular low values both in mass flow and enthalpy have been found in corner subchannel, in disagreement with code predictions, but confirming previous General Electric experiments carried out in a 9-rod test section. This report deals only with the experimental procedure and the results
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.)
Boiling heat transfer on horizontal tube bundles
International Nuclear Information System (INIS)
Anon.
1987-01-01
Nucleate boiling heat transfer characteristics for a tube in a bundle differ from that for a single tube in a pool and this difference is known as 'tube bundle effect.' There exist two bundle effects, positive and negative. The positive bundle effect enhances heat transfer due to convective flow induced by rising bubbles generated from the lower tubes, while the negative bundle effect deteriorates heat transfer due to vapor blanketing caused by accumulation of bubbles. Staggered tube bundles tested and found that the upper tubes in bundles have higher heat transfer coefficients than the lower tubes. The effects of various parameters such as pressure, tube geometry and oil contamination on heat transfer have been examined. Some workers attempted to clarify the mechanism of occurrence of 'bundle effect' by testing tube arrangements of small scale. All reported only enhancement in heat transfer but results showed the symptom of heat transfer deterioration at higher heat fluxes. As mentioned above, it has not been clarified so far even whether the 'tube bundle effect' should serve as enhancement or deterioration of heat transfer in nucleate boiling. In this study, experiments are performed in detail by using bundles of small scale, and effects of heat flux distribution, pressure and tube location are clarified. Furthermore, some consideration on the mechanisms of occurrence of 'tube bundle effect' is made and a method for prediction of heat transfer rate is proposed
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)
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
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.)
Impact of bundle deformation on CHF: ASSERT-PV assessment of extended burnup Bruce B bundle G85159W
International Nuclear Information System (INIS)
Rao, Y.F.; Manzer, A.M.
2005-01-01
This paper presents a subchannel thermalhydraulic analysis of the effect on critical heat flux (CHF) of bundle deformation such as element bow and diametral creep. The bundle geometry is based on the post-irradiation examination (PIE) data of a single bundle from the Bruce B Nuclear Generating Station, Bruce B bundle G85159W, which was irradiated for more than two years in the core during reactor commissioning. The subchannel code ASSERT-PV IST is used to assess changes in CHF and dryout power due to bundle deformation, compared to the reference, undeformed bundle. (author)
Satyanarayana, G.; Narayana, K. L.; Boggarapu, Nageswara Rao
2018-03-01
In the nuclear industry, a critical welding process is joining of an end plate to a fuel rod to form a fuel bundle. Literature on zirconium welding in such a critical operation is limited. A CFD model is developed and performed for the three-dimensional non-linear thermo-fluid analysis incorporating buoyancy and Marnangoni stress and specifying temperature dependent properties to predict weld geometry and temperature field in and around the melt pool of laser spot during welding of a zirconium alloy E110 endplate with a fuel rod. Using this method, it is possible to estimate the weld pool dimensions for the specified laser power and laser-on-time. The temperature profiles will estimate the HAZ and microstructure. The adequacy of generic nature of the model is validated with existing experimental data.
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.)
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
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-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
International Nuclear Information System (INIS)
Yang, T.F.; Wan, A.; Gierszewski, P.; Rapperport, E.; Montgomery, D.B.
1982-01-01
This report presents a preliminary bundle divertor conceptual design for installation on the TEXTOR tokamak. An advanced cascade T-shaped coil configuration is used. This divertor design has the following important characteristics: (1) the current density in the conductor is less than 6 kAmp/cm 2 , and the maximum field is less than 6 Tesla; (2) the divertor can be operated at steady-state either for copper or superconducting conductors; (3) the power consumption is about 7 MW for a normal conductor; (4) the divertor can be inserted into the existing geometry of TEXTOR; (5) the ripple on axis is only 0.3% and the mirror ratio is 2 to 4; (6) the stagnation axis is concave toward the plasma, therefore q/sub D/ is smaller, the acceptance angle is larger, and the efficiency may be better than the conventional circular coil design
International Nuclear Information System (INIS)
Yang, T.F.; Wan, A.; Gierszewski, P.; Rapperport, E.; Montgomery, D.B.
1982-01-01
This report presents a preliminary bundle divertor conceptual design for installation on the TEXTOR tokamak. An advanced cascade T-shaped coil configuration is used. This divertor design has the following important characteristics: (1) the current density in the conductor is less than 6 kAmp/cm 2 , and the maximum field is less than 6 Tesla; (2) the divertor can be operated at steady-state either for copper or superconducting conductors; (3) the power consumption is about 7 MW for a normal conductor; (4) the divertor can be inserted into the existing geometry of TEXTOR; (5) the ripple on axis is only 0.3% and the mirror ratio is 2 to 4; (6) the stagnation axis is concave toward the plasma, therefore q/sub D/ is smaller, the acceptance angle is larger, and the efficiency may be better than the conventional circular coil design
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.
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
International Nuclear Information System (INIS)
Strominger, A.
1990-01-01
A special manifold is an allowed target manifold for the vector multiplets of D=4, N=2 supergravity. These manifolds are of interest for string theory because the moduli spaces of Calabi-Yau threefolds and c=9, (2,2) conformal field theories are special. Previous work has given a local, coordinate-dependent characterization of special geometry. A global description of special geometries is given herein, and their properties are studied. A special manifold M of complex dimension n is characterized by the existence of a holomorphic Sp(2n+2,R)xGL(1,C) vector bundle over M with a nowhere-vanishing holomorphic section Ω. The Kaehler potential on M is the logarithm of the Sp(2n+2,R) invariant norm of Ω. (orig.)
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)
Polyelectrolyte Bundles: Finite size at thermodynamic equilibrium?
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.
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.)
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)
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.)
CFD analysis of liquid metal cooled rod assembly
International Nuclear Information System (INIS)
Son, H.M.; Suh, K.Y.
2007-01-01
The model subassembly of the BREST-type reactor core is a pin bundle of square arrangement. In this bundle there are two zones which differ with respect to pin diameters and level of heat production. The model pin bundle contains one spacer grid which is located near the midplane of the rod bundle geometry. The coolant consists of a eutectic alloy of 22% sodium (Na) plus 78% potassium (K). Experiments were performed in order to observe the thermal hydraulic behavior of the liquid metal coolant in the BREST core simulator. Results were obtained for the coolant exit temperatures, central measuring pin simulator external surface temperatures, and coolant velocities at the perimeter of the measuring pin simulator. A computational fluid dynamics (CFD) code is used to simulate the liquid metal flows in subchannels. Semi-fine mesh structures were used to model the flow with reasonable accuracy and speed once rigorous node resolution dependency had been tested. A subchannel analysis code was used to investigate the flows as well. Since the subchannel analysis code is based on a lumped parameter model, it only calculates the subchannel averaged velocity values. The CFD code results were averaged on the subchannel basis to be comparable with the results from the subchannel code. The mixing vane is not considered for the time being so as to simplify the problem and to reduce the computational cost. The two codes showed similar results. The difference between the experimental and computational results is considered to mainly originate from the existence of the mixing vane. (authors)
CFD analysis of liquid metal cooled rod assembly
Energy Technology Data Exchange (ETDEWEB)
Son, H.M.; Suh, K.Y. [Seoul National Univ. (Korea, Republic of)
2007-07-01
The model subassembly of the BREST-type reactor core is a pin bundle of square arrangement. In this bundle there are two zones which differ with respect to pin diameters and level of heat production. The model pin bundle contains one spacer grid which is located near the midplane of the rod bundle geometry. The coolant consists of a eutectic alloy of 22% sodium (Na) plus 78% potassium (K). Experiments were performed in order to observe the thermal hydraulic behavior of the liquid metal coolant in the BREST core simulator. Results were obtained for the coolant exit temperatures, central measuring pin simulator external surface temperatures, and coolant velocities at the perimeter of the measuring pin simulator. A computational fluid dynamics (CFD) code is used to simulate the liquid metal flows in subchannels. Semi-fine mesh structures were used to model the flow with reasonable accuracy and speed once rigorous node resolution dependency had been tested. A subchannel analysis code was used to investigate the flows as well. Since the subchannel analysis code is based on a lumped parameter model, it only calculates the subchannel averaged velocity values. The CFD code results were averaged on the subchannel basis to be comparable with the results from the subchannel code. The mixing vane is not considered for the time being so as to simplify the problem and to reduce the computational cost. The two codes showed similar results. The difference between the experimental and computational results is considered to mainly originate from the existence of the mixing vane. (authors)
... 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) — ...
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
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
Control rods in LMFBRs: a physics assessment
International Nuclear Information System (INIS)
McFarlane, H.F.; Collins, P.J.
1982-08-01
This physics assessment is based on roughly 300 control rod worth measurements in ZPPR from 1972 to 1981. All ZPPR assemblies simulated mixed-oxide LMFBRs, representing sizes of 350, 700, and 900 MWe. Control rod worth measurements included single rods, various combinations of rods, and Ta and Eu rods. Additional measurements studied variations in B 4 C enrichment, rod interaction effects, variations in rod geometry, neutron streaming in sodium-filled channels, and axial worth profiles. Analyses were done with design-equivalent methods, using ENDF/B Version IV data. Some computations for the sensitivities to approximations in the methods have been included. Comparisons of these analyses with the experiments have allowed the status of control rod physics in the US to be clearly defined
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.)
Moduli of Parabolic Higgs Bundles and Atiyah Algebroids
DEFF Research Database (Denmark)
Logares, Marina; Martens, Johan
2010-01-01
In this paper we study the geometry of the moduli space of (non-strongly) parabolic Higgs bundles over a Riemann surface with marked points. We show that this space possesses a Poisson structure, extending the one on the dual of an Atiyah algebroid over the moduli space of parabolic vector bundle...
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
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
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
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)
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
Polycation induced actin bundles
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...
Polycation induced actin bundles.
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.
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)
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.)
... 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. ...
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.)
Methodology for the study of the boiling crisis in a nuclear fuel bundle
Energy Technology Data Exchange (ETDEWEB)
Crecy, F. de; Juhel, D. [Commissariat a l`Energie Atomique, Grenoble (France)
1995-09-01
The boiling crisis is one of the phenoumena limiting the available power from a nuclear power plant. It has been widely studied for decades, and numerous data, models, correlations or tables are now available in the literature. If we now try to obtain a general view of previous work in this field, we may note that there are several ways of tackling the subject. The mechanistic models try to model the two-phase flow topology and the interaction between different sublayers, and must be validated by comparison with basic experiments, such as DEBORA, where we try to obtain some detailed informations on the two-phase flow pattern in a pure and simple geometry. This allows us to obtain better knowledge of the so-called {open_quotes}intrinsic effect{close_quotes}. These models are not yet acceptable for nuclear use. As the geometry of the rod bundles and grids has a tremendous importance for the Critical Heat Flux (CHF), it is mandatory to have more precise results for a given fuel rod bundle in a restricted range of parameters: this leads to the empirical approach, using empirical CHF predictors (tables, correlations, splines, etc...). One of the key points of such a method is the obtaining local thermohydraulic values, that is to say the evaluation of the so-called {open_quotes}mixing effect{close_quotes}. This is done by a subchannel analysis code or equivalent, which can be qualified on two kinds of experiments: overall flow measurements in a subchannel, such as HYDROMEL in single-phase flow or GRAZIELLA in two-phase flow, or detailed measurements inside a subchannel, such as AGATE. Nevertheless, the final qualification of a specific nuclear fuel, i.e. the synthesis of these mechanistic and empirical approaches, intrinsic and mixing effects, etc..., must be achieved on a global test such as OMEGA. This is the strategy used in France by CEA and its partners FRAMATOME and EdF.
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.).
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...
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.
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.)
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.)
Spinning geometry = Twisted geometry
International Nuclear Information System (INIS)
Freidel, Laurent; Ziprick, Jonathan
2014-01-01
It is well known that the SU(2)-gauge invariant phase space of loop gravity can be represented in terms of twisted geometries. These are piecewise-linear-flat geometries obtained by gluing together polyhedra, but the resulting geometries are not continuous across the faces. Here we show that this phase space can also be represented by continuous, piecewise-flat three-geometries called spinning geometries. These are composed of metric-flat three-cells glued together consistently. The geometry of each cell and the manner in which they are glued is compatible with the choice of fluxes and holonomies. We first remark that the fluxes provide each edge with an angular momentum. By studying the piecewise-flat geometries which minimize edge lengths, we show that these angular momenta can be literally interpreted as the spin of the edges: the geometries of all edges are necessarily helices. We also show that the compatibility of the gluing maps with the holonomy data results in the same conclusion. This shows that a spinning geometry represents a way to glue together the three-cells of a twisted geometry to form a continuous geometry which represents a point in the loop gravity phase space. (paper)
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
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
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.
Gelation And Mechanical Response of Patchy Rods
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.
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
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
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.)
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
Introduction to the theory of fiber bundles and connections I
International Nuclear Information System (INIS)
Socolvsky, M.
1990-01-01
In lectures 1 and 2 we discuss basic concepts of topology and differential geometry: definition of a topological space and of Hausdorff, compact, connected and paracompact spaces; topological groups and actions of groups on spaces; differentiable manifolds, tangent vectors and 1 forms; partitions of unity and Lie groups. In lecture 3 we present the concept of a fiber bundle and discuss vector bundles and principal bundles. The concept of a connection on a smooth vector bundle is defined in lecture 4, together with the associated concepts of curvature and parallel transport; as an illustration we present the Levi-Civita connection on a Riemannian manifold. Finally, in lecture 5 we define connections on principal bundles and present examples with the Lie groups U(1) and SU(2). For reasons of space the present article only includes lectures 1, 2 and 3. Lectures 4 and 5 will be published in a forthcoming paper. (Author)
Bundles over Quantum RealWeighted Projective Spaces
Directory of Open Access Journals (Sweden)
Tomasz Brzeziński
2012-09-01
Full Text Available The algebraic approach to bundles in non-commutative geometry and the definition of quantum real weighted projective spaces are reviewed. Principal U(1-bundles over quantum real weighted projective spaces are constructed. As the spaces in question fall into two separate classes, the negative or odd class that generalises quantum real projective planes and the positive or even class that generalises the quantum disc, so do the constructed principal bundles. In the negative case the principal bundle is proven to be non-trivial and associated projective modules are described. In the positive case the principal bundles turn out to be trivial, and so all the associated modules are free. It is also shown that the circle (coactions on the quantum Seifert manifold that define quantum real weighted projective spaces are almost free.
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
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.
Deformations of the generalised Picard bundle
International Nuclear Information System (INIS)
Biswas, I.; Brambila-Paz, L.; Newstead, P.E.
2004-08-01
Let X be a nonsingular algebraic curve of genus g ≥ 3, and let Mξ denote the moduli space of stable vector bundles of rank n ≥ 2 and degree d with fixed determinant ξ over X such that n and d are coprime. We assume that if g = 3 then n ≥ 4 and if g = 4 then n ≥ 3, and suppose further that n 0 , d 0 are integers such that n 0 ≥ 1 and nd 0 + n 0 d > nn 0 (2g - 2). Let E be a semistable vector bundle over X of rank n 0 and degree d 0 . The generalised Picard bundle W ξ (E) is by definition the vector bundle over M ξ defined by the direct image p M ξ *(U ξ x p X * E) where U ξ is a universal vector bundle over X x M ξ . We obtain an inversion formula allowing us to recover E from W ξ (E) and show that the space of infinitesimal deformations of W ξ (E) is isomorphic to H 1 (X, End(E)). This construction gives a locally complete family of vector bundles over M ξ parametrised by the moduli space M(n 0 ,d 0 ) of stable bundles of rank n 0 and degree d 0 over X. If (n 0 ,d 0 ) = 1 and W ξ (E) is stable for all E is an element of M(n 0 ,d 0 ), the construction determines an isomorphism from M(n 0 ,d 0 ) to a connected component M 0 of a moduli space of stable sheaves over M ξ . This applies in particular when n 0 = 1, in which case M 0 is isomorphic to the Jacobian J of X as a polarised variety. The paper as a whole is a generalisation of results of Kempf and Mukai on Picard bundles over J, and is also related to a paper of Tyurin on the geometry of moduli of vector bundles. (author)
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)
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
International Nuclear Information System (INIS)
Jeong, J. H.; Lee, K. L.
2016-01-01
The wire spacer has important roles to avoid collisions between adjacent rods, to mitigate a vortex induced vibration, and to enhance convective heat transfer by wire spacer induced secondary flow. Many experimental and numerical works has been conducted to understand the thermal-hydraulics of the wire-wrapped fuel bundles. There has been enormous growth in computing capability. Recently, a huge increase of computer power allows to three-dimensional simulation of thermal-hydraulics of wire-wrapped fuel bundles. In this study, the geometry optimization methodology with RANS based in-house CFD (Computational Fluid Dynamics) code has been successfully developed in air condition. In order to apply the developed methodology to fuel assembly, GGI (General Grid Interface) function is developed for in-house CFD code. Furthermore, three-dimensional flow fields calculated with in-house CFD code are compared with those calculated with general purpose commercial CFD solver, CFX. The geometry optimization methodology with RANS based in-house CFD code has been successfully developed in air condition. In order to apply the developed methodology to fuel assembly, GGI function is developed for in-house CFD code as same as CFX. Even though both analyses are conducted with same computational meshes, numerical error due to GGI function locally occurred in only CFX solver around rod surface and boundary region between inner fluid region and outer fluid region.
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.)
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.)
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.
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
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)
Zheng, Fangyang
2002-01-01
The theory of complex manifolds overlaps with several branches of mathematics, including differential geometry, algebraic geometry, several complex variables, global analysis, topology, algebraic number theory, and mathematical physics. Complex manifolds provide a rich class of geometric objects, for example the (common) zero locus of any generic set of complex polynomials is always a complex manifold. Yet complex manifolds behave differently than generic smooth manifolds; they are more coherent and fragile. The rich yet restrictive character of complex manifolds makes them a special and interesting object of study. This book is a self-contained graduate textbook that discusses the differential geometric aspects of complex manifolds. The first part contains standard materials from general topology, differentiable manifolds, and basic Riemannian geometry. The second part discusses complex manifolds and analytic varieties, sheaves and holomorphic vector bundles, and gives a brief account of the surface classifi...
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.)
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
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.
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...
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.)
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
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
Modeling and simulation performance of sucker rod beam pump
Energy Technology Data Exchange (ETDEWEB)
Aditsania, Annisa, E-mail: annisaaditsania@gmail.com [Department of Computational Sciences, Institut Teknologi Bandung (Indonesia); Rahmawati, Silvy Dewi, E-mail: silvyarahmawati@gmail.com; Sukarno, Pudjo, E-mail: psukarno@gmail.com [Department of Petroleum Engineering, Institut Teknologi Bandung (Indonesia); Soewono, Edy, E-mail: esoewono@math.itb.ac.id [Department of Mathematics, Institut Teknologi Bandung (Indonesia)
2015-09-30
Artificial lift is a mechanism to lift hydrocarbon, generally petroleum, from a well to surface. This is used in the case that the natural pressure from the reservoir has significantly decreased. Sucker rod beam pumping is a method of artificial lift. Sucker rod beam pump is modeled in this research as a function of geometry of the surface part, the size of sucker rod string, and fluid properties. Besides its length, sucker rod string also classified into tapered and un-tapered. At the beginning of this research, for easy modeling, the sucker rod string was assumed as un-tapered. The assumption proved non-realistic to use. Therefore, the tapered sucker rod string modeling needs building. The numerical solution of this sucker rod beam pump model is computed using finite difference method. The numerical result shows that the peak of polished rod load for sucker rod beam pump unit C-456-D-256-120, for non-tapered sucker rod string is 38504.2 lb, while for tapered rod string is 25723.3 lb. For that reason, to avoid the sucker rod string breaks due to the overload, the use of tapered sucker rod beam string is suggested in this research.
Modeling and simulation performance of sucker rod beam pump
International Nuclear Information System (INIS)
Aditsania, Annisa; Rahmawati, Silvy Dewi; Sukarno, Pudjo; Soewono, Edy
2015-01-01
Artificial lift is a mechanism to lift hydrocarbon, generally petroleum, from a well to surface. This is used in the case that the natural pressure from the reservoir has significantly decreased. Sucker rod beam pumping is a method of artificial lift. Sucker rod beam pump is modeled in this research as a function of geometry of the surface part, the size of sucker rod string, and fluid properties. Besides its length, sucker rod string also classified into tapered and un-tapered. At the beginning of this research, for easy modeling, the sucker rod string was assumed as un-tapered. The assumption proved non-realistic to use. Therefore, the tapered sucker rod string modeling needs building. The numerical solution of this sucker rod beam pump model is computed using finite difference method. The numerical result shows that the peak of polished rod load for sucker rod beam pump unit C-456-D-256-120, for non-tapered sucker rod string is 38504.2 lb, while for tapered rod string is 25723.3 lb. For that reason, to avoid the sucker rod string breaks due to the overload, the use of tapered sucker rod beam string is suggested in this research
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)
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)
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
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
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
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
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.)
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
Moduli spaces in algebraic geometry
International Nuclear Information System (INIS)
Goettsche, L.
2000-01-01
This volume of the new series of lecture notes of the Abdus Salam International Centre for Theoretical Physics contains the lecture notes of the School on Algebraic Geometry which took place at the Abdus Salam International Centre for Theoretical Physics from 26 July to 13 August 1999. The school consisted of 2 weeks of lecture courses and one week of conference. The topic of the school was moduli spaces. More specifically the lectures were divided into three subtopics: principal bundles on Riemann surfaces, moduli spaces of vector bundles and sheaves on projective varieties, and moduli spaces of curves
International Nuclear Information System (INIS)
Ryu, Eui-Seung; You, Sung-Chang
2014-01-01
As pressure tube diameter creep increase, the coolant flows through some of the interior subchannels of the fuel bundle are reduced and consequently reduces the Critical Heat Flux (CHF). For this reason, Canadian Utilities have performed the project that developing the new fuel design (modified 37-element bundle) to increase critical heat flux. The modified 37-element (37M) bundle has the same overall geometry as the reference 37-element (37R) bundle that is using in the Wolsong units now but the center element diameter has been reduced from 13.06mm to 11.5mm. The reduction in center element diameter of the 37M bundle design increase the flow of center areas to improve the cooling and thus to enhance CHF. The CHF experiments with 37M bundle string simulator in un-crept and crept (3.3%, 5.1% peak creep) flow channels were completed at Stern Laboratories in 2008. A substantially large increase in dryout-power was observed for the 37M bundle compared to the 37R bundle, particularly in the 5.1% crept channel. As a result of the experiments, Ontario Power Generation (OPG) and Bruce Power (BP) have increased the operational margin with this CHF correlation and has fully refueled the 37M fuel on some units or almost done on the other units. KHNP also has performed the project to refuel the 37M bundle which is the same design with OPG and BP recently. This paper summarizes the comparison assessment of Thermalhydraulic (T/H) behavior for 37M bundle and 37R bundle with their own correlations and geometry parameters. This analysis performed with the thermal hydraulic code (NUCIRC) and the site measured data at the Wolsong Unit2. Tests to evaluate the CHF performance with the 37M fuel bundle have been conducted in 2008 using the un-crept, 3.3% crept and 5.1% crept flow channels in the CHF Test facility at Stern Laboratories. In addition pressure drop tests have been performed at the same time. The changes of geometry from 37R bundle to 37M bundle reduced the center element
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
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...
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.)
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
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
Vector bundles on complex projective spaces
Okonek, Christian; Spindler, Heinz
1980-01-01
This expository treatment is based on a survey given by one of the authors at the Séminaire Bourbaki in November 1978 and on a subsequent course held at the University of Göttingen. It is intended to serve as an introduction to the topical question of classification of holomorphic vector bundles on complex projective spaces, and can easily be read by students with a basic knowledge of analytic or algebraic geometry. Short supplementary sections describe more advanced topics, further results, and unsolved problems.
International Nuclear Information System (INIS)
Banas, A.O.; Carver, M.B.; Leung, J.C.H.; Bromley, B.P.
1992-10-01
The general purpose computational fluid dynamics code, Harwell-FLOW3D, has been used to simulate the effects of fuel rod obstructions on pressure drop and heat transfer in single phase turbulent flows in a concentric annular channel. The results of two and three dimensional simulations are reported for obstructions approximating the geometry of bearing pads used in 37 element CANDU fuel bundles. Pressure drop penalty and augmentation of heat transfer have been quantified and correlated with the obstruction geometrical parameters and the dimensionless numbers representing operating conditions. The predicted effects on pressure drop have been compared with several experimental correlations, yielding good agreement. The methodology presented offers results that can be used directly as input into thermalhydraulic analyses in subchannel and system codes. (Author) (23 figs., 15 refs.)
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.)
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.
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)
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 ...
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
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
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.
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.
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)
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
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
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.
Status of thermohydraulic studies of wire-wrapped bundles
International Nuclear Information System (INIS)
Khairallah, A.; Leteinturier, D.; Skok, J.
1979-01-01
A status review is presented of the work undertaken in CEA to acquire good understanding and description of the single-phase thermal-hydraulic problems in LMFBR wire-wrapped bundles. Design-type and reference-type calculational tools developed for the study of forced convection in nominal and distorted bundle geometries are briefly presented. Local hot spots and mixed convection situations are discussed in some more details. Out-of-pile and in-pile experimental programs designed in support to code development are described. (author)
Multi-rod burst test under a loss-of coolant accident condition, (4)
International Nuclear Information System (INIS)
Otomo, Takashi; Hashimoto, Masao; Kawasaki, Satoru; Furuta, Teruo; Uetsuka, Hiroshi
1983-06-01
Multi-rod burst test of No.7808 bundle was performed in steam to estimate quantitative coolant flow channel restriction caused by the ballooning of zircaloy claddings in a fuel assembly during a LOCA transient in LWRs. The test was conducted under the condition that the initial internal pressure in each rod was 35kg/cm 2 (RT) and the heating rate was 9 0 C/s in steam with flow rate of 0.4g/cm 2 .min. The following results were obtained; (1) Maximum and burst pressures in rods were in the range 45 to 48kg/cm 2 and 41 to 45kg/cm 2 , respectively. The burst temperature of cladding were estimated to be 850 to 880 0 C. (2) Axial portions of tubes with greater than 34% strain were observed in the range 0 to 40mm in most rod. The mean length was 19mm in the bundle. (3) The degree of maximum increase in cross-sectional area is 54.2% in the bundle(7 x 7) and 66.9% in the internal rods(5 x 5). (4) Maximum channel area restriction was 40.5% in the bundle(7 x 7) and 51.4% in the internal rods(5 x 5). (author)
Iversen, Birger
1992-01-01
Although it arose from purely theoretical considerations of the underlying axioms of geometry, the work of Einstein and Dirac has demonstrated that hyperbolic geometry is a fundamental aspect of modern physics
Development of input data to energy code for analysis of reactor fuel bundles
International Nuclear Information System (INIS)
Carre, F.O.; Todreas, N.E.
1975-05-01
The ENERGY 1 code is a semi-empirical method for predicting temperature distributions in wire wrapped rod bundles of a LMFBR. A comparison of ENERGY 1 and MISTRAL 2 is presented. The predictions of ENERGY 1 for special sets of data taken under geometric conditions at the limits of the code are analyzed. 14 references
International Nuclear Information System (INIS)
Sadatomi, M.; Kawahara, A.; Sato, Y.
1996-01-01
This paper presents a simple method for predicting the single-phase turbulent mixing rate between adjacent subchannels in nuclear fuel bundles. In this method, the mixing rate is computed as the sum of the two components of turbulent diffusion and convective transfer. Of these, the turbulent diffusion component is calculated using a newly defined subchannel geometry factor F* and the mean turbulent diffusivity for each subchannel which is computed from Elder's equation. The convective transfer component is evaluated from a mixing Stanton number correlation obtained empirically in this study. In order to confirm the validity of the proposed method, experimental data on turbulent mixing rate were obtained using a tracer technique under adiabatic conditions with three test channels, each consisting of two subchannels. The range of Reynolds number covered was 5000-66 000. From comparisons of the predicted turbulent mixing rates with the experimental data of other investigators as well as the authors, it has been confirmed that the proposed method can predict the data in a range of gap clearance to rod diameter ratio of 0.02-0.4 within about ±25% for square array bundles and about ±35% for triangular array bundles. (orig.)
International Nuclear Information System (INIS)
Irie, Tomoo.
1991-01-01
The present invention concerns a control rod in a PWR type reactor. A control rod has an inner cladding tube and an outer cladding tube disposed coaxially, and a water draining hole is formed at the inside of the inner cladding tube. Neutron absorbers are filled in an annular gap between the outer cladding tube and the inner cladding tube. The water draining hole opens at the lower end thereof to the top end of the control rod and at the upper end thereof to the side of the upper end plug of the control rod. If the control rod is dropped to a control rod guide thimble for reactor scram, coolants from the control rod guide thimble are flown from the lower end of the water draining hole and discharged from the upper end passing through the water draining hole. In this way, water from the control rod guide thimble is removed easily when the control rod is dropped. Further, the discharging amount of water itself is reduced by the provision of the water draining hole. Accordingly, sufficient control rod dropping speed can be attained. (I.N.)
Donaldson invariants in algebraic geometry
International Nuclear Information System (INIS)
Goettsche, L.
2000-01-01
In these lectures I want to give an introduction to the relation of Donaldson invariants with algebraic geometry: Donaldson invariants are differentiable invariants of smooth compact 4-manifolds X, defined via moduli spaces of anti-self-dual connections. If X is an algebraic surface, then these moduli spaces can for a suitable choice of the metric be identified with moduli spaces of stable vector bundles on X. This can be used to compute Donaldson invariants via methods of algebraic geometry and has led to a lot of activity on moduli spaces of vector bundles and coherent sheaves on algebraic surfaces. We will first recall the definition of the Donaldson invariants via gauge theory. Then we will show the relation between moduli spaces of anti-self-dual connections and moduli spaces of vector bundles on algebraic surfaces, and how this makes it possible to compute Donaldson invariants via algebraic geometry methods. Finally we concentrate on the case that the number b + of positive eigenvalues of the intersection form on the second homology of the 4-manifold is 1. In this case the Donaldson invariants depend on the metric (or in the algebraic geometric case on the polarization) via a system of walls and chambers. We will study the change of the invariants under wall-crossing, and use this in particular to compute the Donaldson invariants of rational algebraic surfaces. (author)
van den Broek, P.M.
1984-01-01
The aim of this paper is to give a detailed exposition of the relation between the geometry of twistor space and the geometry of Minkowski space. The paper has a didactical purpose; no use has been made of differential geometry and cohomology.
Boiling water reactor fuel bundle
International Nuclear Information System (INIS)
Weitzberg, A.
1986-01-01
A method is described of compensating, without the use of control rods or burnable poisons for power shaping, for reduced moderation of neutrons in an uppermost section of the active core of a boiling water nuclear reactor containing a plurality of elongated fuel rods vertically oriented therein, the fuel rods having nuclear fuel therein, the fuel rods being cooled by water pressurized such that boiling thereof occurs. The method consists of: replacing all of the nuclear fuel in a portion of only the upper half of first predetermined ones of the fuel rods with a solid moderator material of zirconium hydride so that the fuel and the moderator material are axially distributed in the predetermined ones of the fuel rods in an asymmetrical manner relative to a plane through the axial midpoint of each rod and perpendicular to the axis of the rod; placing the moderator material in the first predetermined ones of the fuel rods in respective sealed internal cladding tubes, which are separate from respective external cladding tubes of the first predetermined ones of the fuel rods, to prevent interaction between the moderator material and the external cladding tube of each of the first predetermined ones of the fuel rods; and wherein the number of the first predetermined ones of the fuel rods is at least thirty, and further comprising the steps of: replacing with the moderator material all of the fuel in the upper quarter of each of the at least thirty rods; and also replacing with the moderator material all of the fuel in the adjacent lower quarter of at least sixteen of the at least thirty rods
Bundle Security Protocol for ION
Burleigh, Scott C.; Birrane, Edward J.; Krupiarz, Christopher
2011-01-01
This software implements bundle authentication, conforming to the Delay-Tolerant Networking (DTN) Internet Draft on Bundle Security Protocol (BSP), for the Interplanetary Overlay Network (ION) implementation of DTN. This is the only implementation of BSP that is integrated with ION.
CANFLEX fuel bundle impact test
International Nuclear Information System (INIS)
Chang, Seok Kyu; Chung, C. H.; Park, J. S.; Hong, S. D.; Kim, B. D.
1997-08-01
This document outlines the test results for the impact test of the CANFLEX fuel bundle. Impact test is performed to determine and verify the amount of general bundle shape distortion and defect of the pressure tube that may occur during refuelling. The test specification requires that the fuel bundles and the pressure tube retain their integrities after the impact test under the conservative conditions (10 stationary bundles with 31kg/s flow rate) considering the pressure tube creep. The refuelling simulator operating with pneumatic force and simulated shield plug were fabricated and the velocity/displacement transducer and the high speed camera were also used in this test. The characteristics of the moving bundle (velocity, displacement, impacting force) were measured and analyzed with the impact sensor and the high speed camera system. The important test procedures and measurement results were discussed as follows. 1) Test bundle measurements and the pressure tube inspections 2) Simulated shield plug, outlet flange installation and bundle loading 3) refuelling simulator, inlet flange installation and sensors, high speed camera installation 4) Perform the impact test with operating the refuelling simulator and measure the dynamic characteristics 5) Inspections of the fuel bundles and the pressure tube. (author). 8 refs., 23 tabs., 13 figs
Connections on discrete fibre bundles
International Nuclear Information System (INIS)
Manton, N.S.; Cambridge Univ.
1987-01-01
A new approach to gauge fields on a discrete space-time is proposed, in which the fundamental object is a discrete version of a principal fibre bundle. If the bundle is twisted, the gauge fields are topologically non-trivial automatically. (orig.)
Sasakian and Parabolic Higgs Bundles
Biswas, Indranil; Mj, Mahan
2018-03-01
Let M be a quasi-regular compact connected Sasakian manifold, and let N = M/ S 1 be the base projective variety. We establish an equivalence between the class of Sasakian G-Higgs bundles over M and the class of parabolic (or equivalently, ramified) G-Higgs bundles over the base N.
Twisted Vector Bundles on Pointed Nodal Curves
Indian Academy of Sciences (India)
Abstract. Motivated by the quest for a good compactification of the moduli space of -bundles on a nodal curve we establish a striking relationship between Abramovich's and Vistoli's twisted bundles and Gieseker vector bundles.
International Nuclear Information System (INIS)
Murakami, Kiyonobu; Kurosaki, Akira.
1988-01-01
Purpose: To rapidly insert safety rods for a criticality experiment device into a reactor core container to stop the criticality reaction thereby prevent reactivity accidents. Constitution: A cylinder device having a safety rod as a cylinder rod attached with a piston at one end is constituted. The piston is elevated by pressurized air and attracted and fixed by an electromagnet which is a stationary device disposed at the upper portion of the cylinder. If the current supply to the electromagnet is disconnected, the safety rod constituting the cylinder rod is fallen together with the piston to the lower portion of the cylinder. Since the cylinder rod driving device has neither electrical motor nor driving screw as in the conventional device, necessary space can be reduced and the weight is decreased. In addition, since the inside of the nuclear reactor can easily be shielded completely from the external atmosphere, leakage of radioactive materials can be prevented. (Horiuchi, T.)
Absorbing rods for nuclear fast neutron reactor absorbing assembly
International Nuclear Information System (INIS)
Aji, M.; Ballagny, A.; Haze, R.
1986-01-01
The invention proposes a neutron absorber rod for neutron absorber assembly of a fast neutron reactor. The assembly comprises a bundle of vertical rods, each one comprising a stack of pellets made of a neutron absorber material contained in a long metallic casing with a certain radial play with regard to this casing; this casing includes traps for splinters from the pellets which may appear during reactor operation, at the level of contact between adjacent pellets. The present invention prevents the casing from rupture involved by the disintegration of the pellets producing pieces of boron carbide of high hardness [fr
International Nuclear Information System (INIS)
Bailey, W.J.
1985-04-01
Two of the factors that need to be taken into account with rod consolidation are (1) the effects on rods from their removal from the fuel assembly and (2) the effects on rods as a result of the consolidation process. Potential components of both factors are described in the report. Discussed under (1) are scratches on the fuel rod surfaces, rod breakage, crud, extended burnup, and possible cladding embrittlement due to hydrogen injection at BWRs. Discussed under (2) are the increased water temperature (less than 10 0 C) because of closer packing of the rods, formation of crevices between rods in the close-packed mode, contact with dissimilar metals, and the potential for rapid heating of fuel rods following the loss of water from a spent fuel storage pool. Another factor that plays an important role in rod consolidation is the cost of disposal of the nonfuel-bearing components of the fuel assembly. Also, the dose rate from the components - especially Inconel spacer grids - can affect the handling procedures. Several licensing issues that exist are described. A list of recommendations is provided. 98 refs., 5 figs., 5 tabs
Geometry of the cotangent bundle with Sasakian metrics and its ...
Indian Academy of Sciences (India)
complex structure; Killing vector field. 2010 Mathematics Subject Classification. ... the scalar product g−1 = (gij ) is defined on the cotangent space π−1(x) = T ∗ x Mn by g−1(ω,θ) = gij ωiθj ..... product structure determining the horizontal distribution and the distribution consisting of the tangent planes to fibres [16]. We put.
The burnable poisons utilization for fissile enriched CANDU fuel bundle
Energy Technology Data Exchange (ETDEWEB)
Serghiuta, D; Nainer, O [Team 3 Solutions, Don Mills, ON (Canada)
1996-12-31
Utilization of burnable poison for the fissile enriched fueled CANDU 6 Mk1 core is investigated. The main incentives for this analysis are the reduction of void reactivity effects, the maximization of the fissile content of fresh fuel bundles, and the achievement of better power shape control, in order to preserve the power envelope of the standard 37 rod fuel bundle. The latter allows also the preservation of construction parameters of the standard core (for example: number and location of reactivity devices). It also permits the use of regular shift fueling schemes. The paper makes analyses of MOX weapons-grade plutonium and 1.2% SEU fueled CANDU 6 Mk 1 cores. (author). 6 refs., 4 tabs., 10 figs.
Evaluating big deal journal bundles.
Bergstrom, Theodore C; Courant, Paul N; McAfee, R Preston; Williams, Michael A
2014-07-01
Large commercial publishers sell bundled online subscriptions to their entire list of academic journals at prices significantly lower than the sum of their á la carte prices. Bundle prices differ drastically between institutions, but they are not publicly posted. The data that we have collected enable us to compare the bundle prices charged by commercial publishers with those of nonprofit societies and to examine the types of price discrimination practiced by commercial and nonprofit journal publishers. This information is of interest to economists who study monopolist pricing, librarians interested in making efficient use of library budgets, and scholars who are interested in the availability of the work that they publish.
Control rod studies in small and medium sized fast reactors
International Nuclear Information System (INIS)
John, T.M.; Mohanakrishnan, P.; Mahalakshmi, B.; Singh, R.S.
1988-01-01
Control rods are the primary safety mechanism in the operation of fast reactors. Neutronic parameters associated with the control rods have to be evaluated precisely for studying the behaviour of the reactor under various operating conditions. Control rods are strong neutron absorbers discretely distributed in the reactor core. Accurate estimation of control rod parameters demand, in principle transport theory solutions in exact geometry. But computer codes for such evaluations usually consume exorbitantly large computer time and memory for even a single parameter evaluation. During the design of reactors, evaluation of these parameters will be required for many configurations of control rods. In this paper, the method used at Indira Gandhi Centre for Atomic Research for estimating the parameters associated with control rods is presented. Diffusion theory solutions were used for computations. A scheme using three dimensional geometry represented by triangular meshes and diffusion theory solutions in few energy groups for control rod parameter evaluation is presented. This scheme was employed in estimating the control rod parameters in a 500 Mw(e) fast reactor. Error due to group collapsing is estimated by comparing with 25 group calculations in three dimensions for typical cases. (author). 5 refs, 4 figs, 3 tabs
Fuel bundle to pressure tube fretting in Bruce and Darlington
Energy Technology Data Exchange (ETDEWEB)
Norsworthy, A G; Ditschun, A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)
1996-12-31
As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs.
Fuel bundle to pressure tube fretting in Bruce and Darlington
International Nuclear Information System (INIS)
Norsworthy, A.G.; Ditschun, A.
1995-01-01
As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs
Rodger, Alison
1995-01-01
Molecular Geometry discusses topics relevant to the arrangement of atoms. The book is comprised of seven chapters that tackle several areas of molecular geometry. Chapter 1 reviews the definition and determination of molecular geometry, while Chapter 2 discusses the unified view of stereochemistry and stereochemical changes. Chapter 3 covers the geometry of molecules of second row atoms, and Chapter 4 deals with the main group elements beyond the second row. The book also talks about the complexes of transition metals and f-block elements, and then covers the organometallic compounds and trans
International Nuclear Information System (INIS)
Miyamoto, Yoshiyuki; Higashigawa, Yuichi.
1996-01-01
The present invention provides a control rod terminating system in a BWR type nuclear power plant, which stops an induction electric motor as rapidly as possible to terminate the control rods. Namely, the control rod stopping system controls reactor power by inserting/withdrawing control rods into a reactor by driving them by the induction electric motor. The system is provided with a control device for controlling the control rods and a control device for controlling the braking device. The control device outputs a braking operation signal for actuating the braking device during operation of the control rods to stop the operation of the control rods. Further, the braking device has at least two kinds of breaks, namely, a first and a second brakes. The two kinds of brakes are actuated by receiving the brake operation signals at different timings. The brake device is used also for keeping the control rods after the stopping. Even if a stopping torque of each of the breaks is small, different two kinds of brakes are operated at different timings thereby capable of obtaining a large stopping torque as a total. (I.S.)
International Nuclear Information System (INIS)
Futatsugi, Masao.
1980-01-01
Purpose: To secure the reactor operation safety by the provision of a fluid pressure detecting section for control rod driving fluid and a control rod interlock at the midway of the flow pass for supplying driving fluid to the control rod drives. Constitution: Between a driving line and a direction control valve are provided a pressure detecting portion, an alarm generating device, and a control rod inhibition interlock. The driving fluid from a driving fluid source is discharged by way of a pump and a manual valve into the reactor in which the control rods and reactor fuels are contained. In addition, when the direction control valve is switched and the control rods are inserted and extracted by the control rod drives, the pressure in the driving line is always detected by the pressure detection section, whereby if abnormal pressure is resulted, the alarm generating device is actuated to warn the abnormality and the control rod inhibition interlock is actuated to lock the direction control valve thereby secure the safety operation of the reactor. (Seki, T.)
Indian Academy of Sciences (India)
Bacteria exhibit a wide variety of shapes but the commonly studied species of bacteria are generally either spherical in shape which are called cocci (singular coccus) or have a cylindrical shape and are called rods or bacilli (singular bacillus). In reality rods and cocci are the ends of a continuum. Sonle of the cocci are.
International Nuclear Information System (INIS)
Oonuki, Koji.
1981-01-01
Purpose: To increase the driving speed of control rods at rapid insertion with an elongate control rod and an extension pipe while ensuring sufficient buffering performance in a short buffering distance, by providing a plurality of buffers to an extension pipe between a control rod drive source and a control rod in LMFBR type reactor. Constitution: First, second and third buffers are respectively provided to an acceleration piston, an extension pipe and a control rod respectively and the insertion positions for each of the buffers are displaced orderly from above to below. Upon disconnection of energizing current for an electromagnet, the acceleration piston, the extension pipe and the control rod are rapidly inserted in one body. The first, second and third buffers are respectively actuated at each of their falling strokes upon rapid insertion respectively, and the acceleration piston, the extension pipe and the control rod receive the deceleration effect in the order correspondingly. Although the compression force is applied to the control rod only near the stroke end, it does not cause deformation. (Kawakami, Y.)
Energy Technology Data Exchange (ETDEWEB)
Uchida, Katsuya; Matsuda, Yasuhiko
1984-05-02
The purpose of the project is to enable failed fuel rod detection simply with no requirement for dismantling the fuel assembly. A gamma-ray detection section is arranged so as to attend on the optional fuel rods in the fuel assembly. The fuel assembly is adapted such that a gamma-ray shielding plate is detachably inserted into optional gaps of the fuel rods or, alternatively, the fuel assembly can detachably be inserted to the gamma-ray shielding plate. In this way, amount of gaseous fission products accumulated in all of the plenum portions in the fuel rods as the object of the measurement can be determined without dismantling the fuel assembly. Accordingly, by comparing the amounts of the gaseous fission products, the failed fuel rod can be detected.
International Nuclear Information System (INIS)
Hull, C.M.
1993-01-01
The geometric structure of theories with gauge fields of spins two and higher should involve a higher spin generalisation of Riemannian geometry. Such geometries are discussed and the case of W ∝ -gravity is analysed in detail. While the gauge group for gravity in d dimensions is the diffeomorphism group of the space-time, the gauge group for a certain W-gravity theory (which is W ∝ -gravity in the case d=2) is the group of symplectic diffeomorphisms of the cotangent bundle of the space-time. Gauge transformations for W-gravity gauge fields are given by requiring the invariance of a generalised line element. Densities exist and can be constructed from the line element (generalising √detg μν ) only if d=1 or d=2, so that only for d=1,2 can actions be constructed. These two cases and the corresponding W-gravity actions are considered in detail. In d=2, the gauge group is effectively only a subgroup of the symplectic diffeomorphisms group. Some of the constraints that arise for d=2 are similar to equations arising in the study of self-dual four-dimensional geometries and can be analysed using twistor methods, allowing contact to be made with other formulations of W-gravity. While the twistor transform for self-dual spaces with one Killing vector reduces to a Legendre transform, that for two Killing vectors gives a generalisation of the Legendre transform. (orig.)
MODULAR BUNDLE ADJUSTMENT FOR PHOTOGRAMMETRIC COMPUTATIONS
Directory of Open Access Journals (Sweden)
N. Börlin
2018-05-01
Full Text Available In this paper we investigate how the residuals in bundle adjustment can be split into a composition of simple functions. According to the chain rule, the Jacobian (linearisation of the residual can be formed as a product of the Jacobians of the individual steps. When implemented, this enables a modularisation of the computation of the bundle adjustment residuals and Jacobians where each component has limited responsibility. This enables simple replacement of components to e.g. implement different projection or rotation models by exchanging a module. The technique has previously been used to implement bundle adjustment in the open-source package DBAT (Börlin and Grussenmeyer, 2013 based on the Photogrammetric and Computer Vision interpretations of Brown (1971 lens distortion model. In this paper, we applied the technique to investigate how affine distortions can be used to model the projection of a tilt-shift lens. Two extended distortion models were implemented to test the hypothesis that the ordering of the affine and lens distortion steps can be changed to reduce the size of the residuals of a tilt-shift lens calibration. Results on synthetic data confirm that the ordering of the affine and lens distortion steps matter and is detectable by DBAT. However, when applied to a real camera calibration data set of a tilt-shift lens, no difference between the extended models was seen. This suggests that the tested hypothesis is false and that other effects need to be modelled to better explain the projection. The relatively low implementation effort that was needed to generate the models suggest that the technique can be used to investigate other novel projection models in photogrammetry, including modelling changes in the 3D geometry to better understand the tilt-shift lens.
Modular Bundle Adjustment for Photogrammetric Computations
Börlin, N.; Murtiyoso, A.; Grussenmeyer, P.; Menna, F.; Nocerino, E.
2018-05-01
In this paper we investigate how the residuals in bundle adjustment can be split into a composition of simple functions. According to the chain rule, the Jacobian (linearisation) of the residual can be formed as a product of the Jacobians of the individual steps. When implemented, this enables a modularisation of the computation of the bundle adjustment residuals and Jacobians where each component has limited responsibility. This enables simple replacement of components to e.g. implement different projection or rotation models by exchanging a module. The technique has previously been used to implement bundle adjustment in the open-source package DBAT (Börlin and Grussenmeyer, 2013) based on the Photogrammetric and Computer Vision interpretations of Brown (1971) lens distortion model. In this paper, we applied the technique to investigate how affine distortions can be used to model the projection of a tilt-shift lens. Two extended distortion models were implemented to test the hypothesis that the ordering of the affine and lens distortion steps can be changed to reduce the size of the residuals of a tilt-shift lens calibration. Results on synthetic data confirm that the ordering of the affine and lens distortion steps matter and is detectable by DBAT. However, when applied to a real camera calibration data set of a tilt-shift lens, no difference between the extended models was seen. This suggests that the tested hypothesis is false and that other effects need to be modelled to better explain the projection. The relatively low implementation effort that was needed to generate the models suggest that the technique can be used to investigate other novel projection models in photogrammetry, including modelling changes in the 3D geometry to better understand the tilt-shift lens.