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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-15

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

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)

Flow in rod bundles

International Nuclear Information System (INIS)

Hazi, G.; Mayer, G.

2005-01-01

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

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

International Nuclear Information System (INIS)

Kawahara, Akimaro; Sadatomi, Michio; Higuchi, Tatsuya

2009-01-01

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

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.)

CFD analyses of the rod bowing effect on the subchannel outlet temperature distribution

Energy Technology Data Exchange (ETDEWEB)

Ekstroem, Karoliina; Toppila, Timo [Fortum Power and Heat, Fortum (Finland)

2017-09-15

In the Loviisa 1 and 2 nuclear power plants the subcooling margin of the hottest subchannel of the fuel assembly is monitored. The temperature of the coolant in the hottest subchannel is limited to the constant saturation temperature. Bending of the fuel rods occurs during normal operation due to the differences in the heat profiles of the rods. The coolant temperature will rise more in the subchannel with smaller flow area due to the bending and this has to be taken into account in the safety margin of subchannel enthalpy rise. Computational Fluid Dynamics (CFD) simulations are used to estimate how much the estimated maximum bow of a rod affects the temperature rise of the subchannel. The quantitative uncertainty of the predicted enthalpy rise in fuel bundle subchannel is estimated based on the uncertainty of modelling of mixing between subchannels. The measured turbulence quantities from LDA measurements of cold test assembly made in 1990s in Fortum are compared with CFD results to give uncertainty estimation for turbulence, which is further used for uncertainty estimation of mixing and simulated subchannel enthalpy rise.

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)

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)

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)

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

International Nuclear Information System (INIS)

Rehme, K.

1982-07-01

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

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)

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.

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

International Nuclear Information System (INIS)

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

2005-06-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

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.

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

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

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.

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

International Nuclear Information System (INIS)

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

1977-01-01

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

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

Prediction of gas and liquid turbulent mixing rates between rod bundle subchannels in a two-phase slug-churn flow

International Nuclear Information System (INIS)

Kawahara, Akimaro; Sadatomi, Michio; Tomino, Takayoshi

2000-01-01

This paper presents a slug-churn flow model for predicting turbulent mixing rates of both gas and liquid phases between adjacent subchannels in a BWR fuel rod bundle. In the model, the mixing rate of the liquid phase is calculated as the sum of the three components, i.e., turbulent diffusion, convective transfer and pressure difference fluctuations between the subchannels. The components of turbulent diffusion and convective transfer are calculated from Sadatomi et al.'s (1996) method, applicable to single-phase turbulent mixing, by considering the effect of the increment of liquid velocity due to the presence of gas phase. The component of the pressure difference fluctuations is evaluated from a newly developed correlation. The mixing rate of the gas phase, on the other side, is calculated from a simple relation of mixing rate between gas and liquid phases. The validity of the proposed model has been confirmed with the turbulent mixing rates data of Rudzinski et al. as well as the present authors. (author)

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 subchannel and CFD analysis of void distribution for the BWR fuel bundle test benchmark

International Nuclear Information System (INIS)

In, Wang-Kee; Hwang, Dae-Hyun; Jeong, Jae Jun

2013-01-01

Highlights: ► We analyzed subchannel void distributions using subchannel, system and CFD codes. ► The mean error and standard deviation at steady states were compared. ► The deviation of the CFD simulation was greater than those of the others. ► The large deviation of the CFD prediction is due to interface model uncertainties. -- Abstract: The subchannel grade and microscopic void distributions in the NUPEC (Nuclear Power Engineering Corporation) BFBT (BWR Full-Size Fine-Mesh Bundle Tests) facility have been evaluated with a subchannel analysis code MATRA, a system code MARS and a CFD code CFX-10. Sixteen test series from five different test bundles were selected for the analysis of the steady-state subchannel void distributions. Four test cases for a high burn-up 8 × 8 fuel bundle with a single water rod were simulated using CFX-10 for the microscopic void distribution benchmark. Two transient cases, a turbine trip without a bypass as a typical power transient and a re-circulation pump trip as a flow transient, were also chosen for this analysis. It was found that the steady-state void distributions calculated by both the MATRA and MARS codes coincided well with the measured data in the range of thermodynamic qualities from 5 to 25%. The results of the transient calculations were also similar to each other and very reasonable. The CFD simulation reproduced the overall radial void distribution trend which produces less vapor in the central part of the bundle and more vapor in the periphery. However, the predicted variation of the void distribution inside the subchannels is small, while the measured one is large showing a very high concentration in the center of the subchannels. The variations of the void distribution between the center of the subchannels and the subchannel gap are estimated to be about 5–10% for the CFD prediction and more than 20% for the experiment

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)

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1976-05-01

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

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)

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

Rehme, K.

1983-11-01

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Large eddy simulation of a fuel rod subchannel

International Nuclear Information System (INIS)

Mayer, Gusztav

2007-01-01

In a VVER-440 reactor the measured outlet temperature is related to fuel limit parameters and the power upgrading plans of VVER-440 reactors motivated us to obtain more information on the mixing process of the fuel assemblies. In a VVER-440 rod bundle the fuel rods are arranged in triangular array. Measurement shows (Krauss and Meyer, 1998) that the classical engineering approach, which tries to trace the characterization of such systems back to equivalent (hydraulic diameter) pipe flows, does not give reasonable results. Due to the different turbulence characteristics, the mixing is more intensive in rod bundles than it would be expected based on equivalent pipe flow correlations. As a possible explanation of the high mixing, secondary flow was deduced from measurements by several experimentalists (Trupp and Azad, 1975). Another candidate to explain the high mixing is the so-called flow pulsation phenomenon (Krauss and Meyer, 1998). In this paper we present subchannel simulations (Mayer et al. 2007) using large eddy simulation (LES) methodology and the lattice Boltzmann method (LBM) without the spacers at Reynolds number 21000. The simulation results are compared with the measurements of Trupp and Azad (1975). The mean axial velocity profile shows good agreement with the measurement data. Secondary flow has been observed directly in the simulation results. Reasonable agreement has been achieved for most Reynolds stresses. Nevertheless, the calculated normal stresses show small, but systematic deviation from the measurement data. (author)

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

International Nuclear Information System (INIS)

Rehme, K.

1984-12-01

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

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

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

International Benchmark based on Pressurised Water Reactor Sub-channel and Bundle Tests. Volume III: Departure from Nucleate Boiling

International Nuclear Information System (INIS)

Rubin, Adam; Avramova, Maria; Velazquez-Lozada, Alexander

2016-03-01

This report summarised the second phase of the Nuclear Energy Agency (NEA) and the Nuclear Regulatory Commission (NRC) Benchmark Based on NUPEC PWR Sub-channel and Bundle Tests (PSBT), which was intended to provide data for the verification of Departure from Nucleate Boiling (DNB) prediction in existing thermal-hydraulics codes and provide direction in the development of future methods. This phase was composed of three exercises; Exercise 1: fluid temperature benchmark, Exercise 2: steady-state rod bundle benchmark and Exercise 3: transient rod bundle benchmark. The experimental data provided to the participants of this benchmark is from a series of void measurement tests using full-size mock-up tests for both BWRs and PWRs. These tests were performed from 1987 to 1995 by the Nuclear Power Engineering Corporation (NUPEC) in Japan and made available by the Japan Nuclear Energy Safety Organisation (JNES) for the purposes of this benchmark, which was organised by Pennsylvania State University. Nine institutions from seven countries participated in this benchmark. Nine different computer codes were used in Exercise 1, 2 and 3. Among the computer codes were porous media, sub-channel and systems thermal-hydraulic code. The improvement between FLICA-OVAP (sub-channel) and FLICA (sub-channel) was noticeable. The main difference between the two was that FLICA-OVAP implicitly assigned flow regime based on drift flux, while FLICA assumes single phase flows. In Exercises 2 and 3, the codes were generally able to predict the Departure from Nucleate Boiling (DNB) power as well as the axial location of the onset of DNB (for the steady-state cases) and the time of DNB (for the transient cases). It was noted that the codes that used the Electric-Power-Research- Institute (EPRI) Critical-Heat-Flux (CHF) correlation had the lowest mean error in Exercise 2 for the predicted DNB power

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)

An assessment of thermal behavior of the DUPIC fuel bundle by subchannel analysis

International Nuclear Information System (INIS)

Park, Jee Won.

1997-12-01

Thermal behavior of the standard DUPIC fuel has been assessed. The DUPIC fuel bundle has been modeled for a subchannel analysis using the ASSERT-IV code which was developed by AECL. From the calculated mixture enthalpy, equilibrium quality and void fraction distributions of the DUPIC fuel bundle, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. Based upon the subchannel modeling used in this study, the location of minimum CHFR in the DUPIC fuel bundle has been found to be very similar to that of the standard fuel. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction was found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. Since the transverse interchange model between subchannels is important for the behavior of these variables, it is needed to put more effort in validating the transverse interchange model. For the purpose of investigating influence of thermal-hydraulic parameter variations of the DUPIC fuel bundle, four different values of the channel flow rates were used in the subchannel analysis. The effect of the channel flow reduction on thermal-hydraulic parameters have been presented. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundles in CANDU reactors. (author). 12 refs., 3 tabs., 17 figs

Enthalpy and void distributions in subchannels of PHWR fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Park, J W; Choi, H; Rhee, B W [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1999-12-31

Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)

Enthalpy and void distributions in subchannels of PHWR fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Park, J. W.; Choi, H.; Rhee, B. W. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

Two different types of the CANDU fuel bundles have been modeled for the ASSERT-IV code subchannel analysis. From calculated values of mixture enthalpy and void fraction distribution in the fuel bundles, it is found that net buoyancy effect is pronounced in the central region of the DUPIC fuel bundle when compared with the standard CANDU fuel bundle. It is also found that the central region of the DUPIC fuel bundle can be cooled more efficiently than that of the standard fuel bundle. From the calculated mixture enthalpy distribution at the exit of the fuel channel, it is found that the mixture enthalpy and void fraction can be highest in the peripheral region of the DUPIC fuel bundle. On the other hand, the enthalpy and the void fraction were found to be highest in the central region of the standard CANDU fuel bundle at the exit of the fuel channel. This study shows that the subchannel analysis is very useful in assessing thermal behavior of the fuel bundle that could be used in CANDU reactors. 10 refs., 4 figs., 2 tabs. (Author)

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)

Thermal-hydraulic performance analysis of a subchannel with square and triangle fuel rod arrangements using the entropy generation approach

Institute of Scientific and Technical Information of China (English)

S.Talebi; M.M.Valoujerdi

2017-01-01

The present paper discusses entropy generation in fully developed turbulent flows through a subchannel,arranged in square and triangle arrays.Entropy generation is due to contribution of both heat transfer and pressure drop.Our main objective is to study the effect of key parameters such as spacer grid,fuel rod power distribution,Reynolds number Re,dimensionless heat power ω,lengthto-fuel-diameter ratio λ,and pitch-to-diameter ratio ξ on subchannel entropy generation.The analysis explicitly shows the contribution of heat transfer and pressure drop to the total entropy generation.An analytical formulation is introduced to total entropy generation for situations with uniform and sinusoidal rod power distribution.It is concluded that power distribution affects entropy generation.A smoother power profile leads to less entropy generation.The entropy generation of square rod array bundles is more efficient than that of triangular rod arrays,and spacer grids generate more entropy.

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.

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

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

Assessment of the uncertainties of COBRA sub-channel calculations by using a PWR type rod bundle and the OECD NEA UAM and the PSBT benchmarks data

International Nuclear Information System (INIS)

Panka, I.; Kereszturi, A.

2014-01-01

The assessment of the uncertainties of COBRA-IIIC thermal-hydraulic analyses of rod bundles is performed for a 5-by-5 bundle representing a PWR fuel assembly. In the first part of the paper the modeling uncertainties are evaluated in the term of the uncertainty of the turbulent mixing factor using the OECD NEA/NRC PSBT benchmark data. After that the uncertainties of the COBRA calculations are discussed performing Monte-Carlo type statistical analyses taking into account the modeling uncertainties and other uncertainties prescribed in the OECD NEA UAM benchmark specification. Both steady-state and transient cases are investigated. The target quantities are the uncertainties of the void distribution, the moderator density, the moderator temperature and the DNBR. We will see that - beyond the uncertainties of the geometry and the boundary conditions - it is very important to take into account the modeling uncertainties in case of bundle or sub-channel thermo-hydraulic calculations.

Experimental investigation of effect of spacer on two phase turbulent mixing rate in subchannels of pressure tube type BWR

Energy Technology Data Exchange (ETDEWEB)

Verma, Shashi Kant; Sinha, S.L. [National Institute of Technology, Raipur (India). Mechanical Engineering Dept.; Chandraker, D.K. [Bhabha Atomic Research Centre, Mumbai (India). Reactor Design and Development Group

2017-11-15

Turbulent mixing rate between adjacent subchannels in a two-phase flow has been known to be strongly dependent on the flow pattern. The most important aspect of turbulent motion is that the velocity and pressure at a fixed point do not remain constant with time even in steady state but go through very irregular high frequency fluctuations. These fluctuations influence the diffusion of scalar and vector quantities. The Advanced Heavy Water Reactor (AHWR) is a vertical pressure tube type, heavy water moderated and boiling light water cooled natural circulation based reactor. The fuel bundle of AHWR contains 54 fuel rods set in three concentric rings of 12, 18 and 24 fuel rods. This fuel bundle is divided into number of imaginary interacting flow channel called subchannels. Alteration from single phase to two phase flow situation occurs in reactor rod bundle with raise in power. The two phase flow regimes like bubbly, slug-churn, and annular flow are generally encountered in reactor rod bundle. Prediction of thermal margin of the reactor has necessitated the investigation of turbulent mixing rate of coolant between these subchannels under these flow regimes. Thus, it is fundamental to estimate the effect of spacer grids on turbulent mixing between subchannels of AHWR rod bundle.

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)

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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)

Subchannel friction and flow split factor in hexagonal pin bundles with helical spacers

International Nuclear Information System (INIS)

Bloch, M.

1983-01-01

A hydraulic analysis has been carried on for the edge and the corner subchannel flows, which together with the internal subchannel flow constitute the bundle flow of. Pressure drop, flow and temperature measurements have been made for water flow in edge and corner subchannel flows, which were geometrically separated for measurement purpose. (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.)

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)

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)

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.

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

Relation between medium fluid temperature and centroid subchannel temperatures of a nuclear fuel bundle mock-up

International Nuclear Information System (INIS)

Carvalho Tofani, P. de.

1986-01-01

The subchannel method used in nuclear fuel bundle thermal-hydraulic analysis lies in the statement that subchannel fluid temperatures are taken at mixed mean values. However, the development of mixing correlations and code assessment procedures are, sometimes in the literature, based upon the assumption of identity between lumped and local (subchannel centroid) temperature values. The present paper is concerned with the presentation of an approach for correlating lumped to centroid subchannel temperatures, based upon previously formulated models by the author, applied, applied to a nine heated tube bundle experimental data set. (Author) [pt

Relation between medium fluid temperature and centroid subchannel temperatures of a nuclear fuel bundle mock-up

International Nuclear Information System (INIS)

Carvalho Tofani, P. de.

1986-01-01

The subchannel method used in nuclear fuel bundle thermal-hydraulic analysis lies in the statement that subchannel fluid temperatures are taken at mixed mean values. However, the development of mixing correlations and code assessment procedures are, sometimes in the literature, based upon the assumption of identity between lumped and local (subchannel centroid) temperature values. The present paper is concerned with the presentation of an approach for correlating lumped to centroid subchannel temperatures, based upon previously formulated models by the author, applied to a nine heated tube bundle experimental data set. (Author) [pt

ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

International Nuclear Information System (INIS)

Rao, Y.F.; Cheng, Z.; Waddington, G.M.; Nava-Dominguez, A.

2014-01-01

Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles

ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca; Nava-Dominguez, A., E-mail: navadoma@aecl.ca

2014-08-15

Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles.

Prediction of the single-phase turbulent mixing rate between two parallel subchannels using a subchannel geometry factor

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.)

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

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)

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

Experimental investigation of the enthalpy and mass flow distribution between subchannels in a BWR cluster geometry (PELCO-S)

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

A numerical study of the influence of the void drift model on the predictions of the assert subchannel code

International Nuclear Information System (INIS)

Tye, P.; Teyssedou, A.; Troche, N.; Kiteley, J.

1996-01-01

One of the factors which is important in order to ensure the continued safe operation of nuclear reactors is the ability to accurately predict the 'Critical Heat Flux' (CHF) throughout the rod bundles in the fuel channel. One method currently used by the Canadian nuclear industry to predict the CHF in the fuel bundles of CANDU reactors is to use the ASSERT subchannel code to predict the local thermal-hydraulic conditions prevailing at each axial location in each subchannel in conjunction with appropriate correlations or the CHF look-up table. The successful application of the above methods depends greatly on the ability of ASSERT to accurately predict the local flow conditions throughout the fuel channel. In this paper, full range qualitative verification tests, using the ASSERT subchannel code are presented which show the influence of the void drift model on the predictions of the local subchannel quality. For typical cases using a 7 rod subset of a full 37 element rod bundle taken from the ASSERT validation database, it will be shown that the void drift term can significantly influence the calculated distribution of the quality in the rod bundle. In order to isolate, as much as possible, the influence of the void drift term this first numerical study is carried out with the rod bundle oriented both vertically and horizontally. Subsequently, additional numerical experiments will be presented which show the influence that the void drift model has on the predicted CHF locations. (author)

Steady state subchannel analysis of AHWR fuel cluster

International Nuclear Information System (INIS)

Dasgupta, A.; Chandraker, D.K.; Vijayan, P.K.; Saha, D.

2006-09-01

Subchannel analysis is a technique used to predict the thermal hydraulic behavior of reactor fuel assemblies. The rod cluster is subdivided into a number of parallel interacting flow subchannels. The conservation equations are solved for each of these subchannels, taking into account subchannel interactions. Subchannel analysis of AHWR D-5 fuel cluster has been carried out to determine the variations in thermal hydraulic conditions of coolant and fuel temperatures along the length of the fuel bundle. The hottest regions within the AHWR fuel bundle have been identified. The effect of creep on the fuel performance has also been studied. MCHFR has been calculated using Jansen-Levy correlation. The calculations have been backed by sensitivity analysis for parameters whose values are not known accurately. The sensitivity analysis showed the calculations to have a very low sensitivity to these parameters. Apart from the analysis, the report also includes a brief introduction of a few subchannel codes. A brief description of the equations and solution methodology used in COBRA-IIIC and COBRA-IV-I is also given. (author)

International Benchmark on Pressurised Water Reactor Sub-channel and Bundle Tests. Volume II: Benchmark Results of Phase I: Void Distribution

International Nuclear Information System (INIS)

Rubin, Adam; Avramova, Maria; Velazquez-Lozada, Alexander

2016-03-01

This report summarised the first phase of the Nuclear Energy Agency (NEA) and the US Nuclear Regulatory Commission Benchmark based on NUPEC PWR Sub-channel and Bundle Tests (PSBT), which was intended to provide data for the verification of void distribution models in participants' codes. This phase was composed of four exercises; Exercise 1: steady-state single sub-channel benchmark, Exercise 2: steady-state rod bundle benchmark, Exercise 3: transient rod bundle benchmark and Exercise 4: a pressure drop benchmark. The experimental data provided to the participants of this benchmark is from a series of void measurement tests using full-size mock-up tests for both Boiling Water Reactors (BWRs) and Pressurised Water Reactors (PWRs). These tests were performed from 1987 to 1995 by the Nuclear Power Engineering Corporation (NUPEC) in Japan and made available by the Japan Nuclear Energy Safety Organisation (JNES) for the purposes of this benchmark, which was organised by Pennsylvania State University. Twenty-one institutions from nine countries participated in this benchmark. Seventeen different computer codes were used in Exercises 1, 2, 3 and 4. Among the computer codes were porous media, sub-channel, systems thermal-hydraulic code and Computational Fluid Dynamics (CFD) codes. It was observed that the codes tended to overpredict the thermal equilibrium quality at lower elevations and under predict it at higher elevations. There was also a tendency to overpredict void fraction at lower elevations and underpredict it at high elevations for the bundle test cases. The overprediction of void fraction at low elevations is likely caused by the x-ray densitometer measurement method used. Under sub-cooled boiling conditions, the voids accumulate at heated surfaces (and are therefore not seen in the centre of the sub-channel, where the measurements are being taken), so the experimentally-determined void fractions will be lower than the actual void fraction. Some of the best

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

Science.gov (United States)

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

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

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

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)

Subchannel analysis program for boiling water reactor fuel bundles based on five conservation equations of two-phase flow

International Nuclear Information System (INIS)

Bessho, Y.; Uchikawa, S.

1985-01-01

A subchannel analysis program, MENUETT, is developed for evaluation of thermal-hydraulic characteristics in boiling water reactor fuel bundles. This program is based on five conservation equations of two-phase flow with the drift-flux correlation. The cross flows are calculated separately for liquid and vapor phases from the lateral momentum conservation equation. The effects of turbulent mixing and void drift are accounted for in the program. The conservation equations are implicitly differentiated with the convective terms by the donor-cell method, and are solved iteratively in the axial and lateral directions. Data of the 3 X 3 rod bundle experiments are used for program verification. The lateral distributions of equilibrium quality and mass flow rate at the bundle exit calculated by the program compare satisfactorily with the experimental results

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)

Application of the subchannel analysis code COBRA III C for liquid sodium

International Nuclear Information System (INIS)

Nissen, K.L.

1981-01-01

The subchannel-analysis code COBRA III C was developed to gain knowledge of mass flow and temperature distribution in rod bundles of light water reactors. A comparison of experimental results for the temperature distribution in a 19 rod bundle with calculations done by the computer program shows the capability of COBRA III C to handle liquid sodium cooling. The code needs sodium properties as well as changed correlations for turbulent mixing and heat transfer at the rod. (orig.) [de

CFD modeling of turbulent mixing through vertical pressure tube type boiling water reactor fuel rod bundles with spacer-grids

Science.gov (United States)

Verma, Shashi Kant; Sinha, S. L.; Chandraker, D. K.

2018-05-01

Numerical simulation has been carried out for the study of natural mixing of a Tracer (Passive scalar) to describe the development of turbulent diffusion in an injected sub-channel and, afterwards on, cross-mixing between adjacent sub-channels. In this investigation, post benchmark evaluation of the inter-subchannel mixing was initiated to test the ability of state-of-the-art Computational Fluid Dynamics (CFD) codes to numerically predict the important turbulence parameters downstream of a ring type spacer grid in a rod-bundle. A three-dimensional Computational Fluid Dynamics (CFD) tool (STAR-CCM+) was used to model the single phase flow through a 30° segment or 1/12th of the cross segment of a 54-rod bundle with a ring shaped spacer grid. Polyhedrons were used to discretize the computational domain, along with prismatic cells near the walls, with an overall mesh count of 5.2 M cell volumes. The Reynolds Stress Models (RSM) was tested because of RSM accounts for the turbulence anisotropy, to assess their capability in predicting the velocities as well as mass fraction of potassium nitrate measured in the experiment. In this way, the line probes are located in the different position of subchannels which could be used to characterize the progress of the mixing along the flow direction, and the degree of cross-mixing assessed using the quantity of tracer arriving in the neighbouring sub-channels. The predicted dimensionless mixing scalar along the length, however, was in good agreement with the measurements downstream of spacers.

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

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.

Study of the distributions of flow rate and enthalpy in the sub-channels of a bundle geometry of nuclear reactors in one and two-phase flow

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

Subchannel and bundle friction factors and flow split parameters for laminar transition and turbulent longitudinal flows in wire wrap spaced hexagonal arrays

International Nuclear Information System (INIS)

Hawley, J.T.; Chiu, C.; Todreas, N.E.; Rohsenow, W.M.

1980-01-01

Correlations are presented for subchannel and bundle friction factors and flowsplit parameters for laminar, transition and turbulent longitudinal flows in wire wrap spaced hexagonal arrays. These results are obtained from pressure drop models of flow in individual subchannels. For turbulent flow, an existing pressure drop model for flow in edge subchannels is extended, and the resulting edge subchannel friction factor is identified. Using the expressions for flowsplit parameters and the equal pressure drops assumption, the interior subchannel and bundle friction factors are obtained. For laminar flow, models are developed for pressure drops of individual subchannels. From these models, expressions for the subchannel friction factors are identified and expressions for the flowsplit parameters are derived

Subchannel and bundle friction factors and flowsplit parameters for laminar, transition, and turbulent longitudinal flows in wire-wrap spaced hexagonal arrays. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Hawley, J.T.; Chiu, C.; Rohsenow, W.M.; Todreas, N.E.

1980-08-01

Correlations are presented for subchannel and bundle friction factors and flowsplit parameters for laminar, transition and turbulent longitudinal flows in wire wrap spaced hexagonal arrays. These results are obtained from pressure drop models of flow in individual subchannels. For turbulent flow, an existing pressure drop model for flow in edge subchannels is extended, and the resulting edge subchannel friction factor is identified. Using the expressions for flowsplit parameters and the equal pressured drop assumption, the interior subchannel and bundle friction factors are obtained. For laminar flow, models are developed for pressure drops of individual subchannels. From these models, expressions for the subchannel friction factors are identified and expressions for the flowsplit parameters are derived.

CTF/STAR-CD off-line coupling for simulation of crossflow caused by mixing vane spacers in rod bundles

International Nuclear Information System (INIS)

Avramova, Maria

2011-01-01

Understanding the impact of the spacer grids on the reactor core thermal-hydraulics involves experimental mockup tests, numerical simulations, and development of reliable empirical or semi-empirical models. The state-of-the-art in modeling spacer effects on the thermal-hydraulic performance of the flow in Light Water Reactor (LWR) rod bundles employs numerical experiments by means of Computational Fluid Dynamics (CFD) calculations. The capabilities of the CFD codes are usually being validated against mock-up tests. Once validated, the CFD predictions can be used for improvement and development of more sophisticated models of the subchannel codes. Because of the involved computational cost, CFD codes can not be yet efficiently utilized for full bundle predictions, while advanced subchannel codes are a powerful tool for LWR safety and design analyses. Subchannel analyses are used for whole LWR core evaluations with relatively short CPU times and reasonable computer resources. The objectives of the presented work were to develop, implement, and qualify an innovative spacer grid model utilizing the Computational Fluid Dynamics within a framework of an efficient subchannel analysis tool. A methodology was developed for off-line coupling between the CFD code STAR-CD and the subchannel code CTF. The developed coupling scheme is flexible in axial mesh overlays. It was developed to be easily adapted to any pair of a CFD and a subchannel code. Separate modeling of the spacer grid effects on the diffusive and on the convective processes was implemented and successfully validated against experimental data. (author)

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

International Nuclear Information System (INIS)

Zhu, Xiaojing; Morooka, Shinichi; Oka, Yoshiaki

2014-01-01

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

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

International Nuclear Information System (INIS)

Hwang, Dae Hyun

1994-02-01

the homogeneous mixture model. In view of the result comparing with Katto's experimental data, it is revealed that the present model gives upper and lower bounds of the dryout locations measured in the boiling system with closed bottom end. The model accuracy is improved as the L/D ratio increases. CHF experiment at low velocities including zero inlet flow conditions are conducted for uniformly heated round tube with various flow obstructors positioned at the top of the heated tube. At very low mass velocities or zero inlet flow conditions, where CHF is expected to be governed by the counter-current flow limitation of liquid and vapor, CHF values tend to decrease as the blockage area by the obstructor increases. This effect cannot be distinguished at higher velocities approximately greater than 50 kg/m 2 s. Furthermore it reveals that the effect of the obstruction shape is not significant for low velocity CHF. 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 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 derivatives of CHF with respect to the independent parameters. Three different CHF correlations for tubes (AECL CHF lookup table, Katto correlation, and Biasi correlation) have been examined with uniformly heated bundle CHF data collected from various sources. A limited number of CHF 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

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

International Nuclear Information System (INIS)

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

1976-01-01

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

Study on effects of mixing vane grids on coolant temperature distribution by subchannel analysis

Energy Technology Data Exchange (ETDEWEB)

Mao, H.; Yang, B.W.; Han, B. [Xi' an Jiaotong Univ., Shaanxi (China). Science and Technology Center for Advanced Nuclear Fuel Research

2016-07-15

Mixing vane grids (MVG) have great influence on coolant temperature field in the rod bundle. The MVG could enhance convective heat transfer between the fuel rod wall and the coolant, and promote inter-subchannel mixing at the same time. For the influence of the MVG on convective heat transfer enhancement, many experiments have been done and several correlations have been developed based on the experimental data. However, inter-subchannel mixing promotion caused by the MVG is not well estimated in subchannel analysis because the information of mixing vanes is totally missing in most subchannel codes. This paper analyzes the influence of mixing vanes on coolant temperature distribution using the improved MVG model in subchannel analysis. The coolant temperature distributions with the MVG are analyzed, and the results show that mixing vanes lead to a more uniform temperature distribution. The performances of split vane grids under different power conditions are evaluated. The results are compared with those of spacer grids without mixing vanes and some conclusions are obtained.

Study on effects of mixing vane grids on coolant temperature distribution by subchannel analysis

International Nuclear Information System (INIS)

Mao, H.; Yang, B.W.; Han, B.

2016-01-01

Mixing vane grids (MVG) have great influence on coolant temperature field in the rod bundle. The MVG could enhance convective heat transfer between the fuel rod wall and the coolant, and promote inter-subchannel mixing at the same time. For the influence of the MVG on convective heat transfer enhancement, many experiments have been done and several correlations have been developed based on the experimental data. However, inter-subchannel mixing promotion caused by the MVG is not well estimated in subchannel analysis because the information of mixing vanes is totally missing in most subchannel codes. This paper analyzes the influence of mixing vanes on coolant temperature distribution using the improved MVG model in subchannel analysis. The coolant temperature distributions with the MVG are analyzed, and the results show that mixing vanes lead to a more uniform temperature distribution. The performances of split vane grids under different power conditions are evaluated. The results are compared with those of spacer grids without mixing vanes and some conclusions are obtained.

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

International Nuclear Information System (INIS)

Moeller, R.

1989-12-01

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

Effects of two-phase mixing and void drift models on subchannel void fraction predictions in vertical bundles

Energy Technology Data Exchange (ETDEWEB)

Leung, K.H. [McMaster Univ., Hamilton, Ontario (Canada)], E-mail: leungk4@mcmaster.ca

2009-07-01

The evaluation of the subchannel code ASSERT against the OECD/NEA BFBT benchmark data demonstrated that at low pressures, the void fraction in the corner and side subchannels of a vertical bundle was over-predicted. Preliminary results suggest that this was due to the use of Carlucci's empirical correlation for void drift beyond its applicable range of pressure. Further examination indicates that the choice of the mixing and void drift models has a negligible effect on the error of the subchannel void fraction predictions. A single, isolated subchannel was simulated and results suggest that the root cause behind the over-prediction is inadequate mixing at the sides and corners of the bundle. Increasing the magnitude of the void drift coefficients in Carlucci's model at low pressure was found to improve the overall accuracy of the predictions. A simple correlation relating {omega} to the outlet pressure was found to increase the number of points falling within experimental error by 1.0%. (author)

Effects of two-phase mixing and void drift models on subchannel void fraction predictions in vertical bundles

International Nuclear Information System (INIS)

Leung, K.H.

2009-01-01

The evaluation of the subchannel code ASSERT against the OECD/NEA BFBT benchmark data demonstrated that at low pressures, the void fraction in the corner and side subchannels of a vertical bundle was over-predicted. Preliminary results suggest that this was due to the use of Carlucci's empirical correlation for void drift beyond its applicable range of pressure. Further examination indicates that the choice of the mixing and void drift models has a negligible effect on the error of the subchannel void fraction predictions. A single, isolated subchannel was simulated and results suggest that the root cause behind the over-prediction is inadequate mixing at the sides and corners of the bundle. Increasing the magnitude of the void drift coefficients in Carlucci's model at low pressure was found to improve the overall accuracy of the predictions. A simple correlation relating Ω to the outlet pressure was found to increase the number of points falling within experimental error by 1.0%. (author)

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.

COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and cores

Energy Technology Data Exchange (ETDEWEB)

Wheeler, C.L.; Stewart, C.W.; Cena, R.J.; Rowe, D.S.; Sutey, A.M.

1976-03-01

The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds of subchannels.

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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.)

Implementation of an Experimental Method for Coupled Subchannel Mixing Measurement

International Nuclear Information System (INIS)

Silin, Nicolas; Juanico, Luis; Delmastro, Dario

2003-01-01

In this work the application of a thermal tracing technique to the measurement of thermal turbulent mixing between coupled subchannels is presented.The experiment was carried out on a real scale model with geometry similar to nuclear fuel element rod bundles.Thermal mixing rates were measured for water flows at different Reynolds numbers

Equilibrium quality and mass flux distributions in an adiabatic three-subchannel test section

International Nuclear Information System (INIS)

Yadigaroglu, G.; Maganas, A.

1993-01-01

An experiment was designed to measure the fully-developed quality and mass flux distributions in an adiabatic three-subchannel test section. The three subchannels had the geometrical characteristics of the corner, side, and interior subchannels of a BWR-5 rod bundle. Data collected with Refrigerant-144 at pressures ranging from 7 to 14 bar, simulating operation with water in the range 55 to 103 bar are reported. The average mass flux and quality in the test section were in the ranges 1300 to 1750 kg/m s and -0.03 to 0.25, respectively. The data are analyzed and presented in various forms

Sub-channel analysis of a HPLWR fuel assembly with STAR-CD

International Nuclear Information System (INIS)

Himmel, Steffen R.; Class, Andreas G.; Schulenberg, Thomas; Laurien, Eckart

2008-01-01

Hofmeister et. al. developed a first design proposal for a HPLWR fuel assembly, consisting of a square 7 by 7 fuel pin arrangement within an assembly box and a water box in the centre, replacing 9 fuel rods. Instead of conventional grid spacers, wire wraps are considered due to good coolant mixing and low pressure drop in either flow direction. Within the present work, a novel approach describing the coolant heat up in the sub-channels of such an assembly has been investigated: the commercial software package STAR-CD has been used as a sub-channel code to investigate the thermal-hydraulic performance of such an HPLWR fuel assembly. The aim of the work is to demonstrate that a widely accepted commercial Computational Fluid Dynamics (CFD) code can be used for full rod bundle analysis by applying minor modifications to it. In steady of writing a dedicated code system with numerical solver routines and post-processing tools for sub-channel analyses, the user benefits from the optimized Graphical User Interface (GUI) already provided in STAR-CD. Moreover, a smooth transition to full three-dimensional modeling of the fluid flow inside rod bundles will be possible with the same code system, if considered to be necessary, just by refining the spatial discretization. Steady-state and transient flow regimes can be studied for design as well as reactor safety analysis. As the STAR-CD code uses the Finite Volume Method (FVM) for spatial discretization, the conservation equations for mass, momentum and energy were modified via user-subroutines to obtain the equations known from the usual sub-channel approach. The method will be explained in detail and results will be discussed. (author)

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

International Nuclear Information System (INIS)

Moeller, R.

1989-02-01

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

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

Method of critical power prediction based on film flow model coupled with subchannel analysis

International Nuclear Information System (INIS)

Tomiyama, Akio; Yokomizo, Osamu; Yoshimoto, Yuichiro; Sugawara, Satoshi.

1988-01-01

A new method was developed to predict critical powers for a wide variety of BWR fuel bundle designs. This method couples subchannel analysis with a liquid film flow model, instead of taking the conventional way which couples subchannel analysis with critical heat flux correlations. Flow and quality distributions in a bundle are estimated by the subchannel analysis. Using these distributions, film flow rates along fuel rods are then calculated with the film flow model. Dryout is assumed to occur where one of the film flows disappears. This method is expected to give much better adaptability to variations in geometry, heat flux, flow rate and quality distributions than the conventional methods. In order to verify the method, critical power data under BWR conditions were analyzed. Measured and calculated critical powers agreed to within ±7%. Furthermore critical power data for a tight-latticed bundle obtained by LeTourneau et al. were compared with critical powers calculated by the present method and two conventional methods, CISE correlation and subchannel analysis coupled with the CISE correlation. It was confirmed that the present method can predict critical powers more accurately than the conventional methods. (author)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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.

Equilibrium quality and mass flux distributions in an adiabatic three-subchannel test section

International Nuclear Information System (INIS)

Yadigaroglu, G.; Maganas, A.

1995-01-01

An experiment was designed to measure the fully developed quality and mass flux distributions in an adiabatic three-subchannel test section. The three subchannels had the geometrical characteristics of the corner, side, and interior subchannels of a boiling water reactor (BWR-5) rod bundle. Data collected with Refrigerant-114 at pressures ranging from 7 to 14 bars, simulating operation with water in the range 55 to 103 bars are reported. The average mass flux and quality in the test section were in the ranges 1,300 to 1,750 kg/m 2 · s and -0.03 to 0.25, respectively. The data are analyzed and presented in various forms

A PWR hot-rod model: Relap5/mod3.2.2.{gamma} as a subchannel code

Energy Technology Data Exchange (ETDEWEB)

Kirsten, I.C.; Jones, J.R. [British Energy, Barnwood, Gloucester (United Kingdom); Kimber, G.R. [Atomic Energy Authority Technology, Winfrith, Dorset (United Kingdom); Page, R. [National Nuclear Corp. Ltd., Cheshire (United Kingdom)

2001-07-01

The use of the PWR transient analysis code RELAP5 for detailed assessment of Departure from Nucleate Boiling (DNB) has previously implied coupling it in some way to a subchannel code, either by direct code-to-code coupling or by transferring core boundary conditions to the subchannel code. This paper shows an alternative by using a group of subchannels modelled in RELAP5 to represent a hot rod. The model consists of three parallel channels, each more refined than its neighbour: The first channel represents a quadrant of the core; the second a quadrant on a fuel assembly and the final channel represents a passage adjacent to a single fuel pin. The model is intended for use as part of point kinetics assessments and each channel is assigned a radial form factor designed to conservatively represent the hottest fuel pins in the reactor core. The main outputs from the model are minimum Departure from Nucleate Boiling Ratio (DNBR) and clad oxidation for the hot rod (lead pin). The DNBR results from the hot-rod model are benchmarked against the subchannel code COBRA 3-CP and the results are presented in this paper. Some of the modelling problems that needed to be resolved are also highlighted. (author)

Investigation of velocity distribution in an inner subchannel of wire wrapped fuel pin bundle of sodium-cooled fast reactor

International Nuclear Information System (INIS)

Nishimura, Masahiro; Kamide, Hideki; Ohshima, Hiroyuki; Kobayashi, Jun; Sato, Hiroyuki

2011-01-01

A sodium cooled fast reactor is designed to attain a high burn-up of core fuel in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity via change of flow area in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the detail of flow velocity distribution in a wire wrapped pin bundle. In this study, water experiments were carried out to investigate the detailed velocity distribution in a subchannel of nominal pin geometry as the first step. These basic data are not only useful for understanding of pin bundle thermal hydraulics but also a code validation. A wire-wrapped 3-pin bundle water model was applied to investigate the detailed velocity distribution in the subchannel which is surrounded by 3 pins with wrapping wire. The test section consists of an irregular hexagonal acrylic duct tube and three pins made of fluorinated resin pins which has nearly the same refractive index with that of water and a high light transmission rate. This enables to visualize the central subchannel through the pins. The velocity distribution in the central subchannel with the wrapping wire was measured by PIV (Particle Image Velocimetry) through a side wall of the duct tube. Typical flow velocity conditions in the pin bundle were 0.36m/s (Re=2,700) and 1.6m/s (Re=13,500). Influence of the wrapping wire on the velocity distributions in vertical and horizontal directions was confirmed. A clockwise swirl flow around the wire was found in subchannel. Significant differences were not recognized between the two cases of Re=2,700 and 13,500 concerning flow patterns. (author)

Use of genetic algorithms for optimization of subchannel simulations

International Nuclear Information System (INIS)

Nava Dominguez, A.

2004-01-01

To facilitate the modeling of a rod fuel bundle, the most common used method consist in dividing the complex cross-sectional area in small subsections called subchannels. To close the system equations, a mixture model is used to represent the intersubchannel interactions. These interactions are as follows: diversion cross-flow, turbulent void diffusion, void drift and buoyancy drift. Amongst these mechanisms, the turbulent void diffusion and void drift are frequently modelled using diffusion coefficients. In this work, a novel approach has been employed where an existing subchannel code coupled to a genetic algorithm code which were used to optimize these coefficients. After several numerical simulations, a new objective function based in the principle of minimum dissipated energy was developed. The use of this function in the genetic algorithm coupled to the subchannel code, gave results in good agreement with the experimental data

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

Turbulent mixing between subchannels in a gas-liquid two-phase flow. For the equilibrium flow without net fluid transfer between subchannels

International Nuclear Information System (INIS)

Kawahara, Akimaro; Sadatomi, Michio; Sato, Yoshifusa; Saito, Hidetoshi.

1995-01-01

To provide data necessary for modeling turbulent mixing between subchannels in a nuclear fuel rod bundle, three experiments were made in series for equilibrium two-phase flows, in which net mass exchange does not occur between subchannels for each phase. The first one was the measurement of turbulent mixing rates of both gas and liquid phases by a tracer technique, using air and water as the working fluids. Three kinds of vertical test channels consisting of two subchannels were used. The data have shown that the turbulent mixing rate of each phase in a two-phase flow is strongly dependent on flow regime. So, to see the relation between turbulent mixing and two-phase flow configuration in the subchannels, the second experiment, flow visualization, was made. It was observed in slug and churn flows that a lateral inter-subchannel liquid flow of a large scale is caused by the successive axial transit of large gas bubbles in each subchannel, and the turbulent mixing for the liquid phase is dominated by this lateral flow. To investigate a driving force of such large scale lateral flow, the third experiment, the measurement of an instantaneous pressure differential between the subchannels, was made. The result showed that there is a close relationship between the liquid phase mixing rate and the magnitude of the pressure differential fluctuation. (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.)

CAPRICORN subchannel code for sodium boiling in LMFBR fuel bundles

International Nuclear Information System (INIS)

Padilla, A. Jr.; Smith, D.E.; O'Dell, L.D.

1983-01-01

The CAPRICORN computer code analyzes steady-state and transient, single-phase and boiling problems in LMFBR fuel bundles. CAPRICORN uses the same type of subchannel geometry as the COBRA family of codes and solves a similar system of conservation equations for mass, momentum, and energy. However, CAPRICORN uses a different numerical solution method which allows it to handle the full liquid-to-vapor density change for sodium boiling. Results of the initial comparison with data (the W-1 SLSF pipe rupture experiment) are very promising and provide an optimistic basis for proceeding with further development

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

International Nuclear Information System (INIS)

Yoshida, Hiroyuki; Takase, Kazuyuki; Suzuki, Takayuki

2009-01-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-04-15

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

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

The effect of mixing-vane arrangements in a subchannel turbulent flow

International Nuclear Information System (INIS)

Ikeno, Tsutomu; Murata, Tamotsu; Kajishima, Takeo

2006-01-01

Large eddy simulation (LES) of developed turbulent flows in a rod bundle was carried out for four spacer designs. The mixing-vanes attached at the spacer were inclined at 30degC or 20deg; they were arranged to promote the swirling or convective flow. These arrangements are possible elements to compose an actual rod bundle. Our LES technique with a consistent higher-order immersed boundary method and a one-equation dynamic sub-grid scale model contributed to an efficient treatment of the complex wall configurations of rods and spacers. The computational results reasonably reproduced experimental results for the drag coefficient and the decay rate of swirling flow. The profiles of the axial velocities and the turbulence intensities indicated reasonable trend for the turbulent flow in the rod bundle. The effect of mixing-vane arrangement on the lateral flows was successfully clarified: the cross flow took the longer way on the rod surface than the swirling flow and then was more significantly influenced by momentum diffusion at the no-slip wall. Therefore, the largely inclined mixing-vanes promoted the cross flow only in the neighborhood of the spacer, the swirling flow inside a subchannel could reach farther downstream than the cross flow. (author)

Validation of a Subchannel Analysis Code MATRA Version 1.0

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Seo, Kyung Won; Kwon, Hyouk

2008-10-15

A subchannel analysis code MATRA has been developed for the thermal hydraulic analysis of SMART core. The governing equations and important models were established, and validation calculations have been performed for subchannel flow and enthalpy distributions in rod bundles under steady-state conditions. The governing equations of the MATRA were on the basis of integral balance equation of the two-phase mixture. The effects of non-homogeneous and non-equilibrium states were considered by employing the subcooled boiling model and the phasic slip model. Solution scheme and main structure of the MATRA code, as well as the difference of MATRA and COBRA-IV-I codes, were summarized. Eight different test data sets were employed for the validation of the MATRA code. The collected data consisted of single-phase subchannel flow and temperature distribution data, single-phase inlet flow maldistribution data, single-phase partial flow blockage data, and two-phase subchannel flow and enthalpy distribution data. The prediction accuracy as well as the limitation of the MATRA code was evaluated from this analysis.

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)

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)

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

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

Energy Technology Data Exchange (ETDEWEB)

Kjellstroem, B

1968-12-15

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

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

International Nuclear Information System (INIS)

Misawa, Takeharu; Yoshida, Hiroyuki; Akimoto, Hajime

2008-01-01

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

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)

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)

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

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

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.

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

International Nuclear Information System (INIS)

Mur, J.; Larrauri, D.

1998-07-01

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

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

Development of a subchannel analysis code MATRA (Ver. α)

International Nuclear Information System (INIS)

Yoo, Y. J.; Hwang, D. H.

1998-04-01

A subchannel analysis code MATRA-α, an interim version of MATRA, has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-IV-I. This MATRA code is a thermal-hydraulic analysis code based on the subchannel approach for calculating the enthalpy and flow distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. MATRA-α has been provided with an improved structure, various functions, and models to give the more convenient user environment and to increase the code accuracy, various functions, and models to give the more convenient user environment and to increase the code accuracy. Among them, the pressure drop model has been improved to be applied to non-square-lattice rod arrays, and the lateral transport models between adjacent subchannels have been improved to increase the accuracy in predicting two-phase flow phenomena. Also included in this report are the detailed instructions for input data preparation and for auxiliary pre-processors to serve as a guide to those who want to use MATRA-α. In addition, we compared the predictions of MATRA-α with the experimental data on the flow and enthalpy distribution in three sample rod-bundle cases to evaluate the performance of MATRA-α. All the results revealed that the prediction of MATRA-α were better than those of COBRA-IV-I. (author). 16 refs., 1 tab., 13 figs

Transient subchannel simulation of sodium boiling in a 37 rods bundle with semi implicit and full implicit algorithms

Energy Technology Data Exchange (ETDEWEB)

Azad, Hamed Moslehi; Shirani, A.S. [Shahid Beheshti Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering

2017-07-15

Thermal hydraulic analysis of sodium boiling in fuel assemblies is an important issue in safety of sodium cooled reactors and subchannel method is an efficient approach in transient two phase flow analyses. Almost all of the subchannel codes which use two-fluid model in two phase flow analysis, are based on semi implicit algorithm. With the full implicit method it is possible to use larger time steps. In order to compare the semi implicit algorithm with full implicit algorithm, two transient subchannel numerical programs which one is based on semi implicit algorithm and the other is based on full implicit algorithm have been written in FORTRAN in this work for simulation of transients in sodium cooled Kompakter-Natriumsiede-Kreislauf (KNS) at the former Kernforschungszentrum Karlsruhe (KfK) in Germany.

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-15

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Analysis of two-phase flow inter-subchannel mass and momentum exchanges by the two-fluid model approach

Energy Technology Data Exchange (ETDEWEB)

Ninokata, H. [Tokyo Institute of Technology (Japan); Deguchi, A. [ENO Mathematical Analysis, Tokyo (Japan); Kawahara, A. [Kumamoto Univ., Kumamoto (Japan)

1995-09-01

A new void drift model for the subchannel analysis method is presented for the thermohydraulics calculation of two-phase flows in rod bundles where the flow model uses a two-fluid formulation for the conservation of mass, momentum and energy. A void drift model is constructed based on the experimental data obtained in a geometrically simple inter-connected two circular channel test sections using air-water as working fluids. The void drift force is assumed to be an origin of void drift velocity components of the two-phase cross-flow in a gap area between two adjacent rods and to overcome the momentum exchanges at the phase interface and wall-fluid interface. This void drift force is implemented in the cross flow momentum equations. Computational results have been successfully compared to experimental data available including 3x3 rod bundle data.

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

COBRA - 3C/KFKI: a digital computer program for steady and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements

International Nuclear Information System (INIS)

Vigassy, J.; Kovacs, L.M.

1977-11-01

COBRA-3C/KFKI is a digital computer program for the CDC-3300 computer in FORTRAN language. The program is a revised version of the original COBRA-3C code. The code calculates steady-state and transient flow and enthalpy transport in rod-bundle nuclear fuel elements in both boiling and nonboiling conditions. The mathematical model is formulated by dividing the bundle flow area into flow subchannels that are assumed to contain one-dimensional flow and are coupled to each other by turbulent and diversion crossflow mixing. The program neglects sonic velocity propagation but allows for a temporal and spatial acceleration of the diversion crossflow in the transverse momentum equation. A semiexplicit finite-difference scheme is used to perform a boundary-value solution where the boundary conditions are the inlet enthalpy, inlet flow rate and exit pressure. (D.P.)

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.

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

International Nuclear Information System (INIS)

Zoran Stosic; Vladimir Stevanovic

2005-01-01

is developed based on the void fraction and steam superficial velocity in churn flow, as well as on the liquid film depletion in annular flow. The model is further improved by introducing the criteria for the liquid phase splitting between liquid film and entrained droplets flow at the onset of annular flow. Also, the appropriate model for the prediction of the lift force coefficient is introduced as being important for the prediction of void fraction in the peripheral channels of the rod bundle. The steam and droplets governing equations are discretized in three-dimensional Cartesian co-ordinates, while the liquid film balance equations are solved along the rods direction. The model is numerically solved with the non-stationary control volume SIMPLE type method. The developed advanced model is applied to the prediction of the CHF in the experimental fuel rod assembly of the Japan Atomic Energy Research Institute. Numerical prediction of water and steam mass flux rates and steam void fraction distribution give a complete 3D thermal-hydraulic picture of steam-water flow within the bundle, including two-phase mixture cross flows between neighbouring subchannels. Obtained CHF numerical values are compared with the measured data and the good agreement is obtained. (authors)

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

Numerical simulation of single bubbles rising through subchannels with interface tracking method

International Nuclear Information System (INIS)

Hiroyuki Yoshida; Takuji Nagayoshi; Hidesada Tamai; Tazuyuki Takase; Hajime Akimoto

2005-01-01

Full text of publication follows: Although the sub-channel codes are used for the thermal-hydraulic analysis of fuel bundles in nuclear reactors from the former, many compositions and empirical equations based on experimental results are needed to predict the two-phase flow behavior in details. When there are no experimental data such as the reduced-moderation light water reactor (RMWR) which is studied by the Japan Atomic Energy Research Institute (JAERI), therefore, it is very difficult to obtain highly precise predictions. The RMWR core has remarkably narrow gap spacing between fuel rods (i.e., around 1 mm) which are arranged at a triangular tight-lattice configuration. To evaluate the feasibility and to optimize the thermal design of the RMWR core, a full-scale bundle test is required. However, several systematic full-scale tests are difficult to perform during an initial design phase from economic and temporal reason. Thus, we made a plan to develop a mechanistic BT model to evaluate the effects of the geometry configuration by a two-phase flow numerical simulation. In the plan of the mechanistic BT model development, three dimensional two-phase flow simulation codes with the interface tracking method, the moving particle semi-implicit method and the advanced two-fluid model are developed. In this study, as a part of this model development, detailed two-phase flow simulation code using interface tracking method (named TPFIT) is developed. In this paper, the results of TPFIT code with the advanced interface tracking method applied to single bubbles behavior through subchannels) to verify TPFIT code performance in complicated flow channel as rod bundles. In the simulation, the flow channel is composed of a square duct and four tubes with outside diameters D = 12 mm. The width and height of the duct are 27.2 mm and 192 mm, respectively. In the flow channel, the tubes are used to simulate fuel rods. One center subchannel and four periphery subchannels exist in the

Predictions of Critical Heat Flux Using the ASSERT-PV Subchannel Code for a CANFLEX Variant Bundle

International Nuclear Information System (INIS)

Onder, Ebru Nihan; Leung, Laurence; Kim, Hung; Rao, Yanfei

2009-01-01

The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced CANDU 1 reactor fuel bundle. Based primarily on the CANFLEX 2 fuel bundle, several geometry changes (such as element sizes and pitchcircle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures

A subchannel based annular flow dryout model

International Nuclear Information System (INIS)

Hammouda, Najmeddine; Cheng, Zhong; Rao, Yanfei F.

2016-01-01

Highlights: • A modified annular flow dryout model for subchannel thermalhydraulic analysis. • Implementation of the model in Canadian subchannel code ASSERT-PV. • Assessment of the model against tube CHF experiments. • Assessment of the model against CANDU-bundle CHF experiments. - Abstract: This paper assesses a popular tube-based mechanistic critical heat flux model (Hewitt and Govan’s annular flow model (based on the model of Whalley et al.), and modifies and implements the model for bundle geometries. It describes the results of the ASSERT subchannel code predictions using the modified model, as applied to a single tube and the 28-element, 37-element and 43-element (CANFLEX) CANDU bundles. A quantitative comparison between the model predictions and experimental data indicates good agreement for a wide range of flow conditions. The comparison has resulted in an overall average error of −0.15% and an overall root-mean-square error of 5.46% with tube data representing annular film dryout type critical heat flux, and in an overall average error of −0.9% and an overall RMS error of 9.9% with Stern Laboratories’ CANDU-bundle data.

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

OECD/NRC Benchmark Based on NUPEC PWR Sub-channel and Bundle Test (PSBT). Volume I: Experimental Database and Final Problem Specifications

International Nuclear Information System (INIS)

Rubin, A.; Schoedel, A.; Avramova, M.; Utsuno, H.; Bajorek, S.; Velazquez-Lozada, A.

2012-01-01

The need to refine models for best-estimate calculations, based on good-quality experimental data, has been expressed in many recent meetings in the field of nuclear applications. The needs arising in this respect should not be limited to the currently available macroscopic methods but should be extended to next-generation analysis techniques that focus on more microscopic processes. One of the most valuable databases identified for the thermal-hydraulics modelling was developed by the Nuclear Power Engineering Corporation (NUPEC), Japan, which includes sub-channel void fraction and departure from nucleate boiling (DNB) measurements in a representative Pressurised Water Reactor (PWR) fuel assembly. Part of this database has been made available for this international benchmark activity entitled 'NUPEC PWR Sub-channel and Bundle Tests (PSBT) benchmark'. This international project has been officially approved by the Japanese Ministry of Economy, Trade, and Industry (METI), the US Nuclear Regulatory Commission (NRC) and endorsed by the OECD/NEA. The benchmark team has been organised based on the collaboration between Japan and the USA. A large number of international experts have agreed to participate in this programme. The fine-mesh high-quality sub-channel void fraction and departure from nucleate boiling data encourages advancement in understanding and modelling complex flow behaviour in real bundles. Considering that the present theoretical approach is relatively immature, the benchmark specification is designed so that it will systematically assess and compare the participants' analytical models on the prediction of detailed void distributions and DNB. The development of truly mechanistic models for DNB prediction is currently underway. The benchmark problem includes both macroscopic and microscopic measurement data. In this context, the sub-channel grade void fraction data are regarded as the macroscopic data and the digitised computer graphic images are the

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

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

Measurement and analysis of flow wall shear stress in an interior subchannel of triangular array rods

International Nuclear Information System (INIS)

Fakori-Monazah, M.R.; Todreas, N.E.

1977-08-01

A simulated model of triangular array rods with pitch to diameter ratio of 1.10 (as a test section) and air as the fluid flow was used to study the LMFBR hydraulic parameters. The wall shear stress distribution around the rod periphery, friction factors, static pressure distributions and turbulence intensity corresponding to various Reynolds numbers ranging from 4140 to 36170 in the central subchannel were measured. Various approaches for measurement of wall shear stress were compared. The measurement was performed using the Preston tube technique with the probe outside diameter equal to 0.014 in

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)

Definition of the local fields of velocity, temperature and turbulent characteristics for axial stabilized fluid in arbitrary formed rod bundle assemblies

International Nuclear Information System (INIS)

Sedov, A.A.; Gagin, V.L.

1995-01-01

For the temperature fields in rod clads of experimental assemblies a good agreement have been got with use of prior calculations by subchannel code COBRA-IV-I, from results of which an additional information about δt/δX 3 distribution was taken. The method of definition the local fields of velocity, turbulent kinetic energy, temperature and eddy diffusivities for one-phase axial stabilized fluids in arbitrary formed rod bundle assemblies with invariable upward geometry was developed. According to this model the AGURA code was worked out to calculate local thermal hydraulic problems in combination with temperature fields in fuel rods and constructive elements of fuel assemblies. The method does not use any prior geometric scales and is based only on invariant local flow parameters: turbulent kinetic energy, velocity field deformation tensor and specific work of inner friction. Verification of this method by available experimental data showed a good agreement of calculation data and findings of velocity and t.k.e. fields, when the secondary flows have not a substantial influence to a balance of axial momentum and turbulent kinetic energy. (author)

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

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

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

Laser Doppler measurement and CFD validation in 3 × 3 bundle flow

Energy Technology Data Exchange (ETDEWEB)

Xiong, Jinbiao, E-mail: xiongjinbiao@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University (China); Yu, Yang [School of Nuclear Science and Engineering, Shanghai Jiao Tong University (China); Yu, Nan; Fu, Xiaoliang [State Nuclear Power Software Development Center, National Energy Key Laboratory of Nuclear Power Software (China); Wang, Hongyan [School of Nuclear Science and Engineering, Shanghai Jiao Tong University (China); Cheng, Xu [Karlsruhe Institute of Technology (Germany); Yang, Yanhua [School of Nuclear Science and Engineering, Shanghai Jiao Tong University (China); State Nuclear Power Software Development Center, National Energy Key Laboratory of Nuclear Power Software (China)

2014-04-01

Highlights: • Five-beam LDV is operated in the three-beam mode to measure 3 × 3 bundle flow. • Correlation and FFT techniques are applied to analyze the flow structure. • Large coherent structure is observed in gaps between different subchannels. • The Reynolds stress models predict weak mixing between different subchannels. - Abstract: The five-beam three-component laser Doppler system is operated in the three-beam two-component mode to measure the 3 × 3 bundle flow with simple grid spacer. Experiment has been conducted at Re = 15,200 and 29,900. According to the experiment result, the root mean square (RMS) of axial velocity fluctuation shows large value in the gap and the near-wall region of the edge sub-channel which is induced by the axial velocity gradient. Significant intensity of lateral velocity fluctuation is observed which indicates the strong lateral mixing in a 3 × 3 rod bundle. Through the correlation analysis coherent structures have been observed in the gap region. The spectral analysis shows that the LDV measurement complies to the Komogorov spectrum law, f{sup −5/3}, well. The low-frequency peak spectral density of the axial velocity fluctuation has been observed in the gap region connecting sub-channels with velocity difference. The performance of the SSG model and the baseline Reynolds stress model are investigated based on the experiment result. The models predict higher axial velocity in the interior sub-channel and lower in the edge and corner ones than the experiment result. Large discrepancy between the calculated and measured axial flow velocity is resulted from failure in calculating the strong negative u{sup ′}w{sup ′¯} in the gap region connecting different sub-channels.

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

International Nuclear Information System (INIS)

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

1980-12-01

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

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)

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

International Nuclear Information System (INIS)

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

2006-03-01

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

Use of a genetic algorithm in a subchannel model

International Nuclear Information System (INIS)

Alberto Teyssedou; Armando Nava-Dominguez

2005-01-01

Full text of publication follows: The channel of a nuclear reactor contains the fuel bundles which are made up of fuel elements distributed in a manner that creates a series of interconnected subchannels through which the coolant flows. Subchannel codes are used to determine local flow variables; these codes consider the complex geometry of a nuclear fuel bundle as being divided in simple parallel and interconnected cells called 'subchannels'. Each subchannel is bounded by the solid walls of the fuel rods or by imaginary boundaries placed between adjacent subchannels. In each subchannel the flow is considered as one dimensional, therefore lateral mixing mechanisms between subchannels should be taken into account. These mixing mechanisms are: Diversion cross-flow, Turbulent mixing, Turbulent void diffusion, Void drift and Buoyancy drift; they are implemented as independent contribution terms in a pseudo-vectorial lateral momentum equation. These mixing terms are calculated with correlations that require the use of empirical coefficients. It has been observed, however, that there is no unique set of coefficients and or correlations that can be used to predict a complete range of experimental conditions. To avoid this drawback, in this paper a Genetic Algorithm (GA) was coupled to a subchannel model. The use of a GA in conjunction with an appropriate objective function allows the subchannel model to internally determine the optimal values of the coefficients without user intervention. The subchannel model requires two diffusion coefficients, the drift flux two-phase flow distribution coefficient, C 0 , and a coefficient used to control the lateral pressure losses. The GA algorithm was implemented in order to find the most appropriate values of these four coefficients. Genetic algorithms (GA) are based on the theory of evolution; thus, the GA manipulates a population of individuals (chromosomes) in order to evolve them towards a best adaptation (fitness criterion) to

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

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

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

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)

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)

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

International Nuclear Information System (INIS)

Moeller, R.; Tschoeke, H.

1980-01-01

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

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

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

Constitutive correlations for wire-wrapped subchannel analysis under forced and mixed convection conditions. Part 1

International Nuclear Information System (INIS)

Cheng, S.K.; Todreas, N.E.

1984-08-01

A simple subchannel analysis method based on the ENERGY series of codes, ENERGY-IV, has been established for predicting the temperature field in a single isolated wire-wrapped Liquid Metal Fast Breeder Reactor (LMFBR) subassembly under steady state forced and mixed convection conditions. The ENERGY-IV is a totally empirical code employed for fast running purposes and requires well calibrated lead length averaged input parameters to achieve satisfactory predictions. These input parameters were identified to be the inlet flow split parameters, the subchannel friction factors, the interchannel mixing parameters, the conduction shape factor, and the transverse velocity at the edge gap. Experiments were performed in a 37-pin wire-wrapped rod bundle with a geometry between that of a typical LMFBR fuel subassembly and blanket subassembly for filling the gap in the available data base for the input parameters. The isokinetic extraction method for measuring subchannel velocity, the pitot-static probe for measuring pressure drop, and the salt tracer injection method for estimating the interchannel mixing, were used in these experiments

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

International Nuclear Information System (INIS)

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

2006-11-01

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

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)

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

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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)

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

Subchannel analysis code development for CANDU fuel channel

International Nuclear Information System (INIS)

Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.

1998-07-01

Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 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.

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

Development of two phase turbulent mixing model for subchannel analysis relevant to BWR

International Nuclear Information System (INIS)

Sharma, M.P.; Nayak, A.K.; Kannan, Umasankari

2014-01-01

A two phase flow model is presented, which predicts both liquid and gas phase turbulent mixing rate between adjacent subchannels of reactor rod bundles. The model presented here is for slug churn flow regime, which is dominant as compared to the other regimes like bubbly flow and annular flow regimes, since turbulent mixing rate is the highest in slug churn flow regime. In this paper, we have defined new dimensionless parameters i.e. liquid mixing number and gas mixing number for two phase turbulent mixing. The liquid mixing number is a function of mixture Reynolds number whereas the gas phase mixing number is a function of both mixture Reynolds number and volumetric fraction of gas. The effect of pressure, geometrical influence of subchannel is also included in this model. The present model has been tested against low pressure and temperature air-water and high pressure and temperature steam-water experimental data found that it shows good agreement with available experimental data. (author)

Study on velocity field in a wire wrapped fuel pin bundle of sodium cooled reactor. Detailed velocity distribution in a subchannel

International Nuclear Information System (INIS)

Sato, Hiroyuki; Kobayashi, Jun; Miyakoshi, Hiroyuki; Kamide, Hideki

2009-01-01

A sodium cooled fast reactor is designed to attain a high burn-up core in a feasibility study on commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity via change of flow area in the subassembly and influence the heat removal capability. Therefore, it is of importance to obtain the flow velocity distribution in a wire wrapped pin bundle. A 2.5 times enlarged 7-pin bundle water model was applied to investigate the detailed velocity distribution in an inner subchannel surrounded by 3 pins with wrapping wire. The test section consisted of a hexagonal acrylic duct tube and fluorinated resin pins which had nearly the same refractive index with that of water and a high light transmission rate. The velocity distribution in an inner subchannel with the wrapping wire was measured by PIV (Particle Image Velocimetry) through the front and lateral sides of the duct tube. In the vertical velocity distribution in a narrow space between the pins, the wrapping wire decreased the velocity downstream of the wire and asymmetric flow distribution was formed between the pin and wire. In the horizontal velocity distribution, swirl flow around the wrapping wire was obviously observed. The measured velocity data are useful for code validation of pin bundle thermalhydraulics. (author)

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.

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

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

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

Pressure drop measurements in LMFBR wire wrapped blanket assemblies

International Nuclear Information System (INIS)

Chiu, C.; Hawley, J.; Rohsenow, W.M.; Todreas, N.E.

1977-07-01

In this experiment, measurements of subchannel static pressure for an interior and edge subchannel were taken at two elevations in two wire-wrapped 61-pin bundles. One of the bundles has geometric characteristics of P/D = 1.067 and H/D = 8.0 (4 inch lead length and 0.501 inch rod diameter) and the other bundle has geometric characteristics of P/D = 1.067 and H/D = 4.0 (2 inch lead length and 0.501 inch rod diameter). The bundle average friction factors as well as the local subchannel friction factors for both interior and edge subchannels were determined from the experimental static pressure data. The average subchannel flow rates for both edge and interior subchannels were determined in a separate experiment. Results show that two correlations suggested by Rehme and Novendstern for the bundle average friction factor cannot predict the data within the range of experimental error. The bundle average friction factors for both bundles under test were underestimated by Rehme's correlation and overestimated by Novendstern's correlation. The results of the local subchannel friction factors indicate the effect of the wire lead length is more pronounced in the interior subchannel friction factor than in the edge subchannel friction factor. As the wire wrap lead length decreases, both interior and edge subchannel friction factors increase

Pressure drop measurements in LMFBR wire wrapped blanket assemblies

Energy Technology Data Exchange (ETDEWEB)

Chiu, C.; Hawley, J.; Rohsenow, W.M.; Todreas, N.E.

1977-07-01

In this experiment, measurements of subchannel static pressure for an interior and edge subchannel were taken at two elevations in two wire-wrapped 61-pin bundles. One of the bundles has geometric characteristics of P/D = 1.067 and H/D = 8.0 (4 inch lead length and 0.501 inch rod diameter) and the other bundle has geometric characteristics of P/D = 1.067 and H/D = 4.0 (2 inch lead length and 0.501 inch rod diameter). The bundle average friction factors as well as the local subchannel friction factors for both interior and edge subchannels were determined from the experimental static pressure data. The average subchannel flow rates for both edge and interior subchannels were determined in a separate experiment. Results show that two correlations suggested by Rehme and Novendstern for the bundle average friction factor cannot predict the data within the range of experimental error. The bundle average friction factors for both bundles under test were underestimated by Rehme's correlation and overestimated by Novendstern's correlation. The results of the local subchannel friction factors indicate the effect of the wire lead length is more pronounced in the interior subchannel friction factor than in the edge subchannel friction factor. As the wire wrap lead length decreases, both interior and edge subchannel friction factors increase.

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.

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

International Nuclear Information System (INIS)

Analytis, G.Th.; Luebbesmeyer, D.

1984-04-01

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

Mapping of the lateral flow field in typical subchannels of a support grid with vanes

International Nuclear Information System (INIS)

McClusky, Heather L.; Holloway, Mary V.; Conover, Timothy A.; Beasley, Donald E.; Conner, Michael E.; Smith III, L. David

2003-01-01

Lateral flow fields in four subchannels of a model rod bundle fuel assembly are measured using particle image velocimetry. Vanes (split-vane pairs) are located on the downstream edge of the support grids in the rod bundle fuel assembly and generate swirling flow. Measurements are acquired at a nominal Reynolds number of 28,000 and for seven streamwise locations ranging from 1.4 to 17.0 hydraulic diameters downstream of the grid. The streamwise development of the lateral flow field is divided into two regions based on the lateral flow structure. In Region I, multiple vortices are present in the flow field and vortex interactions occur. Either a single circular vortex or a hairpin shaped flow structure is formed in Region II. Lateral kinetic energy, maximum lateral velocity, centroid of vorticity, radial profiles of azimuthal velocity, and angular momentum are employed as measures of the streamwise development of the lateral flow field. The particle image velocimetry measurements of the present study are compared with laser doppler velocimetry measurements taken for the identical support grids and flow condition. (author)

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)

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.

Validation of the assert subchannel code: Prediction of CHF in standard and non-standard Candu bundle geometries

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

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

Analyses of subchannel velocity distribution for HANARO fuel assembly

International Nuclear Information System (INIS)

Chae, Hee Taek; Han, Gee Yang; Park, Cheol; Lim, In Cheol

1998-10-01

MATRA-h which is a subchannel analysis computer code is used to evaluate the thermal margin of HANARO core. To estimate core thermal margin, accurate prediction of subchannel velocity is very important. The average subchannel velocities of 18 element fuel assembly were obtained from the results of velocity measurement test. To validate the adequacy of the hydraulic model code predictions were compared with the experimental results for the subchannel velocity distribution in 18 element fuel channel. The calculated subchannel velocity distributions in the central channels were larger than those of experiment. On the other hand the subchannel velocities in the outer channels were smaller. It is speculated that the prediction like as above would make CHF value lower because CHF phenomena had been occurred in the outer fuel element in the bundle CHF test of AECL. The prediction for axial pressure distribution coincided with the experimental results well. (author). 9 refs., 9 tabs., 14 figs

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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)

Effects of fluid properties on the cross-flow between subchannels

International Nuclear Information System (INIS)

Azuma, Mie; Hotta, Akitoshi; Shirai, Hiroshi; Ninokata, Hisashi

2004-01-01

This study is one part of the fundamental research on the development of generalized boiling transition analysis methodology applicable to a wide variety of BWR-type fuel bundle geometries. In this study, quantitative identification of the void drift component in cross-flow is conducted by the combination of the Computational Fluid Dynamics (CFD) interface tracking technique and the Multiple Auto Regressive (MAR) method. The numerical model consists of two subchannels with a communication slit. The cross section is modeled in a reference of the Tapucu model, which is extracted from a symmetrical cell pair of actual square lattice fuel rod bundles. An air-water mixture in the slug and churn regimes at atmospheric pressure and room temperature is used as the working fluid. As a result, a concept on extraction of void drift and turbulent mixing components from cross-flow according to experimental data is confirmed by a numerical approach in which pressure is adjusted at each channel outlet using a model of two parallel channels with a communication slit. Criteria to extract components of the turbulent mixing and the void drift components from cross-flow assuming a multi-component mixture are proposed. Effects of surface tension and density ratio of water and air on the cross-flow between subchannels are investigated under conditions to maintain the gas-liquid evolution process based on the proposed criteria. The qualitative evaluation technique with dependency on surface tension and density ratio was proposed. It is expected to provide supplemental information which is difficult to obtain from experiments. (author)

Numerical determination of lateral loss coefficients for subchannel analysis in nuclear fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Sin Kim; Goon-Cherl Park [Seoul National Univ., Seoul (Korea, Republic of)

1995-09-01

An accurate prediction of cross-flow based on detailed knowledge of the velocity field in subchannels of a nuclear fuel assembly is of importance in nuclear fuel performance analysis. In this study, the low-Reynolds number {kappa}-{epsilon} turbulence model has been adopted in two adjacent subchannels with cross-flow. The secondary flow is estimated accurately by the anisotropic algebraic Reynolds stress model. This model was numerically calculated by the finite element method and has been verified successfully through comparison with existing experimental data. Finally, with the numerical analysis of the velocity field in such subchannel domain, an analytical correlation of the lateral loss coefficient is obtained to predict the cross-flow rate in subchannel analysis codes. The correlation is expressed as a function of the ratio of the lateral flow velocity to the donor subchannel axial velocity, recipient channel Reynolds number and pitch-to-diameter.

Validation of the ASSERT subchannel code for prediction of CHF in standard and non-standard CANDU bundle geometries

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

Validation of the ASSERT subchannel code: Prediction of critical heat flux in standard and nonstandard CANDU bundle geometries

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

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

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

International Nuclear Information System (INIS)

Shen Xiuzhong; Yu Pingan; Yang Guanyue

1996-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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.)

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

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

Recent developments in ASSERT-PV code for subchannel thermalhydraulics

International Nuclear Information System (INIS)

Rao, Y.F.; Hammouda, N.

2003-01-01

This paper summarises recent development of ASSERT-PV, and provides examples of applications to CANDU fuel bundles in predicting flow, heat transfer and sheath temperatures. The development work is intended to improve computational and phenomenological modelling capabilities of ASSERT-PV in simulating various flow scenarios in CANDU fuel bundles. The latest version of ASSERT-PV can be used for simulations of steady state or transient, subchannel thermalhydraulics in CANDU bundles under conditions up to and including post-dryout heat transfer. (author)

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)

Subchannel analysis in nuclear reactors

International Nuclear Information System (INIS)

Ninokata, H.; Aritomi, M.

1992-01-01

This book contains 10 informative papers, presented at the International Seminar on Subchannel Analysis 1992 (ISSCA '92), organized by the Institute of Applied Energy, in collaboration with Atomic Energy Society of Japan, Tokyo Electric Power Company, Kansai Electric Power Company, Nuclear Power Engineering Corporation and the Japan Atomic Energy Research Institute, and held at the TIS-Green Forum, Tokyo, Japan, 30 October 1992. The seminar ISSCA '92 was intended to review the current state-of-the-arts of the method being applied to advanced nuclear reactors including Advanced BWRs, Advanced PWRs and LMRs, and to identify the problems to be solved, improvements to be made, and the needs of R and Ds that were required from the new fuel bundles design. The critical review was to focus on the performances of currently available subchannel analysis codes with regard to heat transfer and fluid flows in various types of nuclear reactor bundles under both steady-state and transient operating conditions, CHF, boiling transition (BT) or dryout behaviors and post BT. The behaviors of physical modeling and numerical methods in these extreme conditions were discussed and the methods critically evaluated in comparison with experiments. (author) (J.P.N.)

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)

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

Assessment of subchannel code ASSERT-PV for flow-distribution predictions

International Nuclear Information System (INIS)

Nava-Dominguez, A.; Rao, Y.F.; Waddington, G.M.

2014-01-01

Highlights: • Assessment of the subchannel code ASSERT-PV 3.2 for the prediction of flow distribution. • Open literature and in-house experimental data to quantify ASSERT-PV predictions. • Model changes assessed against vertical and horizontal flow experiments. • Improvement of flow-distribution predictions under CANDU-relevant conditions. - Abstract: This paper reports an assessment of the recently released subchannel code ASSERT-PV 3.2 for the prediction of flow-distribution in fuel bundles, including subchannel void fraction, quality and mass fluxes. Experimental data from open literature and from in-house tests are used to assess the flow-distribution models in ASSERT-PV 3.2. The prediction statistics using the recommended model set of ASSERT-PV 3.2 are compared to those from previous code versions. Separate-effects sensitivity studies are performed to quantify the contribution of each flow-distribution model change or enhancement to the improvement in flow-distribution prediction. The assessment demonstrates significant improvement in the prediction of flow-distribution in horizontal fuel channels containing CANDU bundles

Assessment of subchannel code ASSERT-PV for flow-distribution predictions

Energy Technology Data Exchange (ETDEWEB)

Nava-Dominguez, A., E-mail: navadoma@aecl.ca; Rao, Y.F., E-mail: raoy@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

2014-08-15

Highlights: • Assessment of the subchannel code ASSERT-PV 3.2 for the prediction of flow distribution. • Open literature and in-house experimental data to quantify ASSERT-PV predictions. • Model changes assessed against vertical and horizontal flow experiments. • Improvement of flow-distribution predictions under CANDU-relevant conditions. - Abstract: This paper reports an assessment of the recently released subchannel code ASSERT-PV 3.2 for the prediction of flow-distribution in fuel bundles, including subchannel void fraction, quality and mass fluxes. Experimental data from open literature and from in-house tests are used to assess the flow-distribution models in ASSERT-PV 3.2. The prediction statistics using the recommended model set of ASSERT-PV 3.2 are compared to those from previous code versions. Separate-effects sensitivity studies are performed to quantify the contribution of each flow-distribution model change or enhancement to the improvement in flow-distribution prediction. The assessment demonstrates significant improvement in the prediction of flow-distribution in horizontal fuel channels containing CANDU bundles.

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)

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

International Nuclear Information System (INIS)

Sengpiel, W.

1980-12-01

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

Experimental investigation of the enthalpy- and mass flow-distribution in 16-rod clusters with BWR-PWR-geometries and conditions

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

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)

Coarse-grid-CFD. An advantageous alternative to subchannel analysis

International Nuclear Information System (INIS)

Class, A.G.; Himmel, S.R.; Viellieber, M.O.

2011-01-01

In the 1960 th to 80 th when current GEN II reactor technology was developed, the only possible approach was to use one-dimensional subchannel analysis to compute the flow inside a fuel bundle so that the subchannel scale could be resolved. For simulations of the whole reactor core either system codes or homogenization were employed. In system codes resolution of individual assemblies was the state of the art. Homogenization used porous media equations simulations and averaged the thermohydraulics on reactor core scale. Current potent computing power allows using Computational Fluid Dynamics (CFD) to simulate individual fuel assemblies. Yet the large number of fuel assemblies within the core forbids exploiting CFD for core wide simulation. We propose to combine ideas of subchannel analysis and CFD to develop a new methodology which takes advantage of the fast development of commercial CFD software and the efficiency of subchannel analysis. In this methodology was first applied to simulate a wire-wrap fuel bundle of the High Performance Light Water Reactor (HPLWR). Computations using an inviscid Euler solver on an extremely coarse grid were tuned to predict the true thermohydraulics by adding volumetric forces. These forces represent the non-resolved sub-grid physics. The volumetric forces cannot be measured directly. However, they can be accessed from detailed CFD simulations resolving all relevant physics. Parameterization of these subgrid forces can be realized analogous to models in subchannel codes. In the present work we extend the methodology to the open source solver OpenFOAM and a specific hexagonal fuel assembly which is studied in the framework of liquid metal cooled GEN IV reactor concepts. (orig.)

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

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

International Nuclear Information System (INIS)

Su Jian; Silva Freire, Atila P.

2002-01-01

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

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)

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

Prediction of turbulent mixing rates of both gas and liquid phases between adjacent subchannels in a two-phase slug-churn flow

International Nuclear Information System (INIS)

Kawahara, A.; Sadatomi, M.; Tomino, T.; Sato, Y.

1998-01-01

This paper presents a slug-churn flow model for predicting turbulent mixing rates of both gas and liquid phase 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 compenents 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 correlations. 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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

The effect of the advanced drift-flux model of ASSERT-PV on critical heat flux, flow and void distributions in CANDU bundle subchannels

International Nuclear Information System (INIS)

Hammouda, N.; Rao, Y.F.

2017-01-01

Highlights: • Presentation of the “advanced” drift-flux model of the subchannel code ASSERT-PV. • Study the effect of the drift-flux model of ASSERT on CHF and flow distribution. • Quantify model component effects with flow, quality and dryout power measurements. - Abstract: This paper studies the effect of the drift flux model of the subchannel code ASSERT-PV on critical heat flux (CHF), void fraction and flow distribution across fuel bundles. Numerical experiments and comparison against measurements were performed to examine the trends and relative behaviour of the different components of the model under various flow conditions. The drift flux model of ASSERT-PV is composed of three components: (a) the lateral component or diversion cross-flow, caused by pressure difference between connected subchannels, (b) the turbulent diffusion component or the turbulent mixing through gaps of subchannels, caused by instantaneous turbulent fluctuations or flow oscillations, and (c) the void drift component that occurs due to the two-phase tendency toward a preferred distribution. This study shows that the drift flux model has a significant impact on CHF, void fraction and flow distribution predictions. The lateral component of the drift flux model has a stronger effect on CHF predictions than the axial component, especially for horizontal flow. Predictions of CHF, void fraction and flow distributions are most sensitive to the turbulent diffusion component of the model, followed by the void drift component. Buoyancy drift can be significant, but it does not have as much influence on CHF and flow distribution as the turbulent diffusion and void drift.

Analyses of HANARO bundle experiment data using MATRA-h: revision

Energy Technology Data Exchange (ETDEWEB)

Lim, In Cheol; Park, Cheol; Chae, Hee Taek; Lee, Choong Sung

1999-08-01

When the construction and operation license for HANARO was renewed in 1995, imposed was a condition that the safety limit CHFR should have the margin of 25 percent. The reason for this were that the number of bundle CHF experiment data was not enough for the validation of the prediction of CHF in bundle geometry and that the ability of COBRA/KMRR to prediction the local coolant condition was not fully validated. For the resolution of this imposition, more bundle CHF data were gathered and the subchannel exit temperature distribution was obtained during the in-core irradiation test of instrumented bundle (Type-B bundle). also, for these experimental data, subchannel analyses were performed by using MATRA-h code which is the modified version of MATRA-a which is a modified version of KAERI's MATRA-a for the application to HANARO. By comparing the analysis results with the experimental results, it was found that the HANARO subchannel analysis method would give the conservative or best-estimated predictions for the CHF in bundle geometry. This report is the revision of KAERI/TR-1090/98 on the analysis of bundle experiment data using MATRA-h. (Author). 16 refs., 16 tabs., 25 figs.

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

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.

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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)

CFD analysis of the dynamic behaviour of a fuel rod subchannel in a supercritical water reactor with point kinetics

International Nuclear Information System (INIS)

Ampomah-Amoako, Emmanuel; Akaho, Edward H.K.; Nyarko, Benjamin J.B.; Ambrosini, Walter

2013-01-01

Highlights: • The analysis of flow stability of nuclear fuel subchannels with supercritical water is presented. • The results obtained by a CFD code are compared with those of a system code. • The model includes also heat conduction in the fuel rod and point neutron kinetics. - Abstract: The paper presents the analysis by a CFD code of coupled neutronic–thermal hydraulic instabilities in a subchannel slice belonging to a square lattice assembly. The work represents a further phase in the assessment of the suitability of CFD codes for studies of flow stability of supercritical fluids; the research started in previous work with the analysis of bare 2D circular pipes and already addressed 3D subchannel slices with no allowance for heat conduction or neutronic effects. In the present phase, a more realistic system is considered, dealing with a slice of a fuel assembly subchannel containing the regions of the pellet, the gap and the cladding and including also the effect of inlet and outlet throttling. The adopted neutronic model is a point kinetics one, including six delayed neutron groups with global Doppler and fluid density feedbacks. The response of the model to perturbations applied starting from a steady-state condition at the rated power is compared with that of a similar model developed for a 1D system code. Upward, horizontal and downward flow orientations are addressed making use of a uniform power profile and changing relevant parameters as the gap equivalent conductance and the density reactivity coefficient. A bottom-peaked power profile is also considered in the case of vertical upward flow. Though the adopted model can still be considered simple in comparison with actual details of fuel assemblies, the obtained results demonstrate the potential of the adopted methodology for more accurate analyses to be made with larger computational resources

Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

International Nuclear Information System (INIS)

Hwang, Dae Hyun; Seo, K. W.

2006-01-01

A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC

Development and Application of Subchannel Analysis Code Technology for Advanced Reactor Systems

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Seo, K. W

2006-01-15

A study has been performed for the development and assessment of a subchannel analysis code which is purposed to be used for the analysis of advanced reactor conditions with various configurations of reactor core and several kinds of reactor coolant fluids. The subchannel analysis code was developed on the basis of MATRA code which is being developed at KAERI. A GUI (Graphic User Interface) system was adopted in order to reduce input error and to enhance user convenience. The subchannel code was complemented in the property calculation modules by including various fluids such as heavy liquid metal, gas, refrigerant,and supercritical water. The subchannel code was applied to calculate the local thermal hydraulic conditions inside the non-square test bundles which was employed for the analysis of CHF. The applicability of the subchannel code was evaluated for a high temperature gas cooled reactor condition and supercritical pressure conditions with water and Freon. A subchannel analysis has been conducted for European ADS(Accelerator-Driven subcritical System) with Pb-Bi coolant through the international cooperation work between KAERI and FZK, Germany. In addition, the prediction capability of the subchannel code was evaluated for the subchannel void distribution data by participating an international code benchmark program which was organized by OECD/NRC.

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

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.

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

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

CANGEO program description and user's instruction

International Nuclear Information System (INIS)

Park, Joo Hwan; Jun, J. S.; Suk, H. C.; Hwang, D. H.; Yoo, Y. J.

1997-10-01

This is a report for the description and user's instruction of CANGEO (CANdu GEOmetry) program. CANGEO program is able to generate the geometry input for thermalhydraulic subchannel analysis of CANDU fuel channel. It is developed for the calculation of subchannel analysis of CANDU fuel channel. It is developed for the calculation of subchannel data for a complicated geometry such as a CANDU fuel channel. And this code can be applicable to a bundle eccentricity from pressure tube center to bundle center, pressure tube creep, rod ballooning, and any symmetric subchannel geometry required by thermalhydraulic subchannel analysis code. The report also describes the basic calculation method of subchannel geometry, including user's manual, input and output for a CANDU fuel channel loaded with 37-element fuel bundle. (author). 35 refs., 6 figs

Analysis of Void Fraction Distribution and Departure from Nucleate Boiling in Single Subchannel and Bundle Geometries Using Subchannel, System, and Computational Fluid Dynamics Codes

Directory of Open Access Journals (Sweden)

Taewan Kim

2012-01-01

Full Text Available In order to assess the accuracy and validity of subchannel, system, and computational fluid dynamics codes, the Paul Scherrer Institut has participated in the OECD/NRC PSBT benchmark with the thermal-hydraulic system code TRACE5.0 developed by US NRC, the subchannel code FLICA4 developed by CEA, and the computational fluid dynamic code STAR-CD developed by CD-adapco. The PSBT benchmark consists of a series of void distribution exercises and departure from nucleate boiling exercises. The results reveal that the prediction by the subchannel code FLICA4 agrees with the experimental data reasonably well in both steady-state and transient conditions. The analyses of single-subchannel experiments by means of the computational fluid dynamic code STAR-CD with the CD-adapco boiling model indicate that the prediction of the void fraction has no significant discrepancy from the experiments. The analyses with TRACE point out the necessity to perform additional assessment of the subcooled boiling model and bulk condensation model of TRACE.

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.

Subchannel analysis of a critical power test, using simulated BWR 8x8 fuel assembly

International Nuclear Information System (INIS)

Mitsutake, T.; Terasaka, H.; Yoshimura, K.; Oishi, M.; Inoue, A.; Akiyama, M.

1990-01-01

Critical power predictions have been compared with the critical power test data obtained in simulated BWR 8x8 fuel rod assemblies. Two analytical methods for the critical power prediction in rod assemblies are used in the prediction, which are the subchannel analysis using the COBRA/BWR subchannel computer code with empirical critical heat flux (CHF) correlations and the liquid film dryout estimation using the CRIPP-3F 'multi-fluid' computer code. Improvements in both the analytical methods were made for spacer effect modeling, though they were specific for application to the current BWR rod assembly type. In general a reasonable agreement was obtained, though comparisons, between the prediction and the obtained test data. (orig.)

A coupling model for the two-stage core calculation method with subchannel analysis for boiling water reactors

International Nuclear Information System (INIS)

Mitsuyasu, Takeshi; Aoyama, Motoo; Yamamoto, Akio

2017-01-01

Highlights: • A coupling model of the two-stage core calculation with subchannel analysis. • BWR fuel assembly parameters are assumed and verified. • The model was evaluated for heterogeneous problems. - Abstract: The two-stage core analysis method is widely used for BWR core analysis. The purpose of this study is to develop a core analysis model coupled with subchannel analysis within the two-stage calculation scheme using an assembly-based thermal-hydraulics calculation in the core analysis. The model changes the 2D lattice physics scheme, and couples with 3D subchannel analysis which evaluates the thermal-hydraulics characteristics within the coolant flow area divided as some subchannel regions. In order to couple with these two analyses, some BWR fuel assembly parameters are assumed and verified. The developed model is evaluated for the heterogeneous problem with and without a control rod. The present model is especially effective for the control rod inserted condition. The present model can incorporate the subchannel effect into the current two-stage core calculation method.

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

International Nuclear Information System (INIS)

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

1986-11-01

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

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)

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

International Nuclear Information System (INIS)

Hassan, Y.A.

1987-01-01

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

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

International Nuclear Information System (INIS)

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

1980-10-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-15

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

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

International Nuclear Information System (INIS)

Nazifi, M.; Nematollahi, M.

2007-01-01

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

MIT extraction method for measuring average subchannel axial velocities in reactor assemblies

International Nuclear Information System (INIS)

Hawley, J.T.; Chiu, C.; Todreas, N.E.

1980-08-01

The MIT extraction method for obtaining flow split data for individual subchannels is described in detail. An analysis of the method is presented which shows that isokinetic values of the subchannel flow rates are obtained directly even though the method is non-isokinetic. Time saving methods are discussed for obtaining the average value of the interior region flow split parameter. An analysis of the method at low bundle flow rates indicates that there is no inherent low flow rate limitation on the method and suggests a way to obtain laminar flow split data

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

International Nuclear Information System (INIS)

1987-01-01

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

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

International Nuclear Information System (INIS)

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

1984-07-01

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

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

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

International Nuclear Information System (INIS)

Silva Junior, H.C. da.

1978-12-01

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

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.)

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.

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)

Numerical simulation of secondary flow in bubbly turbulent flow in sub-channel

International Nuclear Information System (INIS)

Ikeno, Tsutomu; Kataoka, Isao

2009-01-01

Secondary flow in bubbly turbulent flow in sub-channel was simulated by using an algebraic turbulence stress model. The mass, momentum, turbulence energy and bubble diffusion equations were used as fundamental equation. The basis for these equations was the two-fluid model: the equation of liquid phase was picked up from the equation system theoretically derived for the gas-liquid two-fluid turbulent flow. The fundamental equation was transformed onto a generalized coordinate system fitted to the computational domain in sub-channel. It was discretized for the SIMPLE algorism using the finite-volume method. The shape of sub-channel causes a distortion of the computational mesh, and orthogonal nature of the mesh is sometimes broken. An iterative method to satisfy a requirement for the contra-variant velocity was introduced to represent accurate symmetric boundary condition. Two-phase flow at a steady state was simulated for different magnitude of secondary flow and void fraction. The secondary flow enhanced the momentum transport in sub-channel and accelerated the liquid phase in the rod gap. This effect was slightly mitigated when the void fraction increased. The acceleration can contribute to effective cooling in the rod gap. The numerical result implied a phenomenon of industrial interest. This suggested that experimental approach is necessary to validate the numerical model and to identify the phenomenon. (author)

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)

Experiments on the fluid dynamics and thermodynamics of rod bundles to verify and support the design of SNR-300 fuel elements - status and open problems

International Nuclear Information System (INIS)

Moeller, R.; Weinberg, D.; Trippe, G.; Tschoeke, H.

1978-01-01

The reliable design of reactor core elements calls for precise knowledge of the 3D-temperature fields of the different components; this primarily applies to the fuel element cladding tubes, these being the first safety barrier. This paper describes and discusses where and how the 3D-temperature fields so far determined exclusively with the help of global thermohydraulic computer codes (SUBCHANNEL-Codes) have to be determined more accurately by local investigations. The basis of these investigations is the measurement of local velocities and temperatures in 19-rod bundle models of the SNR-300 fuel element performed at the Kernforschungszentrum Karlsruhe (KfK). Some important results of the extensive experimental investigations are reported and compared with global and local recalculations. Open problems are pointed out. The influence of the uncertainties in the thermohydraulic design with respect to the strength analysis are discussed. The most significant results and conclusions are: (1) The peripheral bundle region is the critical zone, which has to be investigated with priority. Here the maximal azimuthal temperature differences of the claddings are ten times higher than those in the central bundle region. (2) The present deviations between thermal experiments and global as well as local calculations are much too high. Within the parameters investigated a careful code adaptation to the experiments is of high priority. (3) The knowledge gaps concerning liquid metal heat transfer in irregular geometries have to be closed. (4) The hot-channel analysis has to be checked with respect to the latest more detailed knowledge of thermohydraulics. (author)

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

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

Investigation on flow patterns and transition characteristics in a tube-bundle channel

International Nuclear Information System (INIS)

Xiang Wenyuan; Lu Yonghong; Zhao Guisheng

2012-01-01

Tube-bundle channels have been widely used in condenser-evaporator and other industrial heat-exchange equipment. The characteristics of two-phase flow patterns and their transitions for refrigerant R-113 through a vertical tube-bundle channel are experimentally investigated using high-speed camera. Experiments show that there are four main flow patterns in the tube-bundle channel, which are bubbly flow, bubbly-churn flow, churn flow and annular flow. And in the same cross-section of tube- bundle channels, it is shown that there might be different flow patterns in different sub-channels. The flow pattern transitions exhibit unsynchronized in different sub-channels. On the basis of experimental research, the flow pattern map is drawn and analyses are made on the comparison of differences between boiling flow patterns in a circular tube and those in a tube-bundle channel. (authors)

Modeling approach for annular-fuel elements using the ASSERT-PV subchannel code

International Nuclear Information System (INIS)

Dominguez, A.N.; Rao, Y.

2012-01-01

The internally and externally cooled annular fuel (hereafter called annular fuel) is under consideration for a new high burn-up fuel bundle design in Atomic Energy of Canada Limited (AECL) for its current, and its Generation IV reactor. An assessment of different options to model a bundle fuelled with annular fuel elements is presented. Two options are discussed: 1) Modify the subchannel code ASSERT-PV to handle multiple types of elements in the same bundle, and 2) coupling ASSERT-PV with an external application. Based on this assessment, the selected option is to couple ASSERT-PV with the thermalhydraulic system code CATHENA. (author)

NCEL: two dimensional finite element code for steady-state temperature distribution in seven rod-bundle

International Nuclear Information System (INIS)

Hrehor, M.

1979-01-01

The paper deals with an application of the finite element method to the heat transfer study in seven-pin models of LMFBR fuel subassembly. The developed code NCEL solves two-dimensional steady state heat conduction equation in the whole subassembly model cross-section and enebles to perform the analysis of thermal behaviour in both normal and accidental operational conditions as eccentricity of the central rod or full or partial (porous) blockage of some part of the cross-flow area. The heat removal is simulated by heat sinks in coolant under conditions of subchannels slug flow approximation

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)

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

International Nuclear Information System (INIS)

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

1983-12-01

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

Vibration of fuel bundles

International Nuclear Information System (INIS)

Chen, S.S.

1975-06-01

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

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

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

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

International Nuclear Information System (INIS)

Larrauri, D.; Briere, E.

1997-12-01

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

Velocity distribution measurement in wire-spaced fuel pin bundle

International Nuclear Information System (INIS)

Mizuta, Hiroshi; Ohtake, Toshihide; Uruwashi, Shinichi; Takahashi, Keiichi

1974-01-01

Flow distribution measurement was made in the subchannels of a pin bundle in air flow. The present paper is interim because the target of this work is the decision of temperature of the pin surface in contact with wire spacers. The wire-spaced fuel pin bundle used for the experiment consists of 37 simulated fuel pins of stainless steel tubes, 3000 mm in length and 31.6 mm in diameter, which are wound spirally with 6 mm stainless steel wire. The bundle is wrapped with a hexagonal tube, 3500 mm in length and 293 mm in flat-to-flat distance. The bundle is fixed with knock-bar at the entrance of air flow in the hexagonal tube. The pitch of pins in the bundle is 37.6 mm (P/D=1.19) and the wrapping pitch of wire is 1100 mm (H/D=34.8). A pair of arrow-type 5-hole Pitot tubes are used to measure the flow velocity and the direction of air flow in the pin bundle. The measurement of flow distribution was made with the conditions of air flow rate of 0.33 m 3 /sec, air temperature of 45 0 C, and average Reynolds number of 15100 (average air velocity of 20.6 m/sec.). It was found that circular flow existed in the down stream of wire spacers, that axial flow velocity was slower in the subchannels, which contained wire spacers, than in those not affected by the wire, and that the flow angle to the axial velocity at the boundary of subchannels was two thirds smaller than wire wrapping angle. (Tai, I.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-15

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

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

International Nuclear Information System (INIS)

Debbarma, Ajoy; Pandey, Krishna Murari

2016-01-01

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

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

International Nuclear Information System (INIS)

Ishida, Naoyuki; Utsuno, Hideaki; Kasahara, Fumio

2003-01-01

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

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.

Input modelling of ASSERT-PV V2R8M1 for RUFIC fuel bundle

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Suk, Ho Chun

2001-02-01

This report describes the input modelling for subchannel analysis of CANFLEX-RU (RUFIC) fuel bundle which has been developed for an advanced fuel bundle of CANDU-6 reactor, using ASSERT-PV V2R8M1 code. Execution file of ASSERT-PV V2R8M1 code was recently transferred from AECL under JRDC agreement between KAERI and AECL. SSERT-PV V2R8M1 which is quite different from COBRA-IV-i code has been developed for thermalhydraulic analysis of CANDU-6 fuel channel by subchannel analysis method and updated so that 43-element CANDU fuel geometry can be applied. Hence, ASSERT code can be applied to the subchannel analysis of RUFIC fuel bundle. The present report was prepared for ASSERT input modelling of RUFIC fuel bundle. Since the ASSERT results highly depend on user's input modelling, the calculation results may be quite different among the user's input models. The objective of the present report is the preparation of detail description of the background information for input data and gives credibility of the calculation results.

Input modelling of ASSERT-PV V2R8M1 for RUFIC fuel bundle

International Nuclear Information System (INIS)

Park, Joo Hwan; Suk, Ho Chun

2001-02-01

This report describes the input modelling for subchannel analysis of CANFLEX-RU (RUFIC) fuel bundle which has been developed for an advanced fuel bundle of CANDU-6 reactor, using ASSERT-PV V2R8M1 code. Execution file of ASSERT-PV V2R8M1 code was recently transferred from AECL under JRDC agreement between KAERI and AECL. SSERT-PV V2R8M1 which is quite different from COBRA-IV-i code has been developed for thermalhydraulic analysis of CANDU-6 fuel channel by subchannel analysis method and updated so that 43-element CANDU fuel geometry can be applied. Hence, ASSERT code can be applied to the subchannel analysis of RUFIC fuel bundle. The present report was prepared for ASSERT input modelling of RUFIC fuel bundle. Since the ASSERT results highly depend on user's input modelling, the calculation results may be quite different among the user's input models. The objective of the present report is the preparation of detail description of the background information for input data and gives credibility of the calculation results

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.)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

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.)

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

International Nuclear Information System (INIS)

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

1987-05-01

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

Development and assessment of a sub-channel code applicable for trans-critical transient of SCWR

International Nuclear Information System (INIS)

Liu, X.J.; Yang, T.; Cheng, X.

2013-01-01

Highlights: • A new sub-channel code COBRA-SC for SCWR is developed. • Pseudo two-phase method is employed to realize trans-critical transient calculation. • Good suitability of COBRA-SC is demonstrated by preliminary assessment. • The calculation results of COBRA-SC agree well with ATHLET code. -- Abstract: In the last few years, extensive R and D activities have been launched covering various aspects of supercritical water-cooled reactor (SCWR), especially the thermal-hydraulic analysis. Sub-channel code plays an indispensable role to predict the detail thermal-hydraulic behavior of the SCWR fuel assembly. This paper develops a new version of sub-channel code COBRA-SC based on the previous COBRA-IV code. The supercritical water property and heat transfer/pressure drop correlations under supercritical pressure are implemented to this code. Moreover, in order to simulate the trans-critical transient (the pressure undergo a decrease from the supercritical pressure to the subcritical pressure), pseudo two-phase method is employed in COBRA-SC code. This work is completed by introduction of a virtual two-phase region near the pseudo-critical line. A smooth transition of void fraction can be realized. In addition, several heat transfer correlations right underneath the critical point are introduced into this code to capture the heat transfer behavior during the trans-critical transient. Some experimental data from simple geometry, e.g. the single tube, small rod bundle, is used to validate and evaluate this new developed COBRA-SC code. The predicted results show a good agreement with the experimental data, demonstrating good feasibility of this code for SCWR condition. A code to code comparison between COBRA-SC and ATHLET for a blowdown transient of a small fuel assembly is also presented and discussed in this paper

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

International Nuclear Information System (INIS)

Fernandez y Fernandez, E.; Carajilescov, P.

1980-11-01

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

Cobra-IE Evaluation by Simulation of the NUPEC BWR Full-Size Fine-Mesh Bundle Test (BFBT)

International Nuclear Information System (INIS)

Burns, C. J.; Aumiler, D.L.

2006-01-01

The COBRA-IE computer code is a thermal-hydraulic subchannel analysis program capable of simulating phenomena present in both PWRs and BWRs. As part of ongoing COBRA-IE assessment efforts, the code has been evaluated against experimental data from the NUPEC BWR Full-Size Fine-Mesh Bundle Tests (BFBT). The BFBT experiments utilized an 8 x 8 rod bundle to simulate BWR operating conditions and power profiles, providing an excellent database for investigation of the capabilities of the code. Benchmarks performed included steady-state and transient void distribution, single-phase and two-phase pressure drop, and steady-state and transient critical power measurements. COBRA-IE effectively captured the trends seen in the experimental data with acceptable prediction error. Future sensitivity studies are planned to investigate the effects of enabling and/or modifying optional code models dealing with void drift, turbulent mixing, rewetting, and CHF

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

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)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Implementation of a phenomenological DNB prediction model based on macroscale boiling flow processes in PWR fuel bundles

International Nuclear Information System (INIS)

Mohitpour, Maryam; Jahanfarnia, Gholamreza; Shams, Mehrzad

2014-01-01

Highlights: • A numerical framework was developed to mechanistically predict DNB in PWR bundles. • The DNB evaluation module was incorporated into the two-phase flow solver module. • Three-dimensional two-fluid model was the basis of two-phase flow solver module. • Liquid sublayer dryout model was adapted as CHF-triggering mechanism in DNB module. • Ability of DNB modeling approach was studied based on PSBT DNB tests in rod bundle. - Abstract: In this study, a numerical framework, comprising of a two-phase flow subchannel solver module and a Departure from Nucleate Boiling (DNB) evaluation module, was developed to mechanistically predict DNB in rod bundles of Pressurized Water Reactor (PWR). In this regard, the liquid sublayer dryout model was adapted as the Critical Heat Flux (CHF) triggering mechanism to reduce the dependency of the model on empirical correlations in the DNB evaluation module. To predict local flow boiling processes, a three-dimensional two-fluid formalism coupled with heat conduction was selected as the basic tool for the development of the two-phase flow subchannel analysis solver. Evaluation of the DNB modeling approach was performed against OECD/NRC NUPEC PWR Bundle tests (PSBT Benchmark) which supplied an extensive database for the development of truly mechanistic and consistent models for boiling transition and CHF. The results of the analyses demonstrated the need for additional assessment of the subcooled boiling model and the bulk condensation model implemented in the two-phase flow solver module. The proposed model slightly under-predicts the DNB power in comparison with the ones obtained from steady-state benchmark measurements. However, this prediction is acceptable compared with other codes. Another point about the DNB prediction model is that it has a conservative behavior. Examination of the axial and radial position of the first detected DNB using code-to-code comparisons on the basis of PSBT data indicated that the our

Input modelling of ASSERT-PV V2R8M1 for RUFIC fuel bundle

Energy Technology Data Exchange (ETDEWEB)

Park, Joo Hwan; Suk, Ho Chun

2001-02-01

This report describes the input modelling for subchannel analysis of CANFLEX-RU (RUFIC) fuel bundle which has been developed for an advanced fuel bundle of CANDU-6 reactor, using ASSERT-PV V2R8M1 code. Execution file of ASSERT-PV V2R8M1 code was recently transferred from AECL under JRDC agreement between KAERI and AECL. SSERT-PV V2R8M1 which is quite different from COBRA-IV-i code has been developed for thermalhydraulic analysis of CANDU-6 fuel channel by subchannel analysis method and updated so that 43-element CANDU fuel geometry can be applied. Hence, ASSERT code can be applied to the subchannel analysis of RUFIC fuel bundle. The present report was prepared for ASSERT input modelling of RUFIC fuel bundle. Since the ASSERT results highly depend on user's input modelling, the calculation results may be quite different among the user's input models. The objective of the present report is the preparation of detail description of the background information for input data and gives credibility of the calculation results.

Inter-subchannel heat transfer modeling for a subchannel analysis of liquid metal-cooled reactors

International Nuclear Information System (INIS)

Hae-Yong, Jeong; Kwi-Seok, Ha; Young-Min, Kwon; Yong-Bum, Lee; Dohee, Hahn

2007-01-01

In a subchannel approach, the temperature, pressure and velocity in a subchannel are averaged, and one representative thermal-hydraulic condition specifies the state of a subchannel. To enhance the predictability of a subchannel analysis code, it is required to model the inter-subchannel heat transfer between the adjacent subchannels as accurately as possible. One of the critical parameters which determine the thermal-hydraulic behavior of the coolant in subchannels is the heat conduction between two neighboring sub-channels. This portion of a heat transfer becomes more important in the design of an LMR (Liquid Metal-cooled Reactor) because of the high heat capacity of the liquid metal coolant. The other important part of heat transfer is the mixing of flow as a form of cross flow. Especially, the turbulent mixing caused by the eddy motion of fluid across the gap between the subchannels enhances the exchange of the momentum and the energy through the gap with no net transport of the mass. Major results of recent efforts on these modeling have been implemented in a subchannel analysis code MATRA-LMR-FB. The analysis shows that the accuracy of a subchannel analysis code is improved by enhancing the models describing the conduction heat transfer and the cross-flow mixing, especially at low flow rate. (authors)

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

International Nuclear Information System (INIS)

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

1978-06-01

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

Validation uncertainty of MATRA code for subchannel void distributions

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae-Hyun; Kim, S. J.; Kwon, H.; Seo, K. W. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-10-15

To extend code capability to the whole core subchannel analysis, pre-conditioned Krylov matrix solvers such as BiCGSTAB and GMRES are implemented in MATRA code as well as parallel computing algorithms using MPI and OPENMP. It is coded by fortran 90, and has some user friendly features such as graphic user interface. MATRA code was approved by Korean regulation body for design calculation of integral-type PWR named SMART. The major role subchannel code is to evaluate core thermal margin through the hot channel analysis and uncertainty evaluation for CHF predictions. In addition, it is potentially used for the best estimation of core thermal hydraulic field by incorporating into multiphysics and/or multi-scale code systems. In this study we examined a validation process for the subchannel code MATRA specifically in the prediction of subchannel void distributions. The primary objective of validation is to estimate a range within which the simulation modeling error lies. The experimental data for subchannel void distributions at steady state and transient conditions was provided on the framework of OECD/NEA UAM benchmark program. The validation uncertainty of MATRA code was evaluated for a specific experimental condition by comparing the simulation result and experimental data. A validation process should be preceded by code and solution verification. However, quantification of verification uncertainty was not addressed in this study. The validation uncertainty of the MATRA code for predicting subchannel void distribution was evaluated for a single data point of void fraction measurement at a 5x5 PWR test bundle on the framework of OECD UAM benchmark program. The validation standard uncertainties were evaluated as 4.2%, 3.9%, and 2.8% with the Monte-Carlo approach at the axial levels of 2216 mm, 2669 mm, and 3177 mm, respectively. The sensitivity coefficient approach revealed similar results of uncertainties but did not account for the nonlinear effects on the

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

CERN Document Server

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

1999-01-01

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

Local heat transfer where heated rods touch in axially flowing water

International Nuclear Information System (INIS)

Kast, S.J.

1983-05-01

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

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

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

Validation of the ASSERT subchannel code for MAPLE-X10 reactor conditions

International Nuclear Information System (INIS)

Carver, M.B.; Kiteley, J.C.; Junop, S.V.; Wasilewicz, J.F.

1993-01-01

The ASSERT subchannel analysis code has been developed specifically to model flow and phase distributions within CANDU fuel channels. Recently, ASSERT has been adapted for use in simulating the MAPLE-X10 reactor. ASSERT uses an advanced drift-flux model, which permits the phases to have unequal velocities and unequal temperatures (UVUT), and thus can model non-equilibrium effects such as phase separation tendencies and subcooled boiling. Modelling subcooled boiling accurately is particularly important for MAPLE-X10. This paper briefly summarizes the non-equilibrium model used in the ASSERT code, the equations used to represent these models, and the algorithms used to solve the equations numerically. Very few modifications to the ASSERT models were needed to address MAPLE conditions. These centered on the manner in which finned fuel rods are treated, and they are discussed in the paper. The paper also gives results from validation exercises, in which the ASSERT code predictions of subcooled boiling void-fraction and critical heat flux were compared to experiments using MAPLE-X10 finned fuel elements in annuli and various bundles. 18 refs., 13 figs., 3 tabs

The drift flux model in the ASSERT subchannel code

International Nuclear Information System (INIS)

Carver, M.B.; Judd, R.A.; Kiteley, J.C.; Tahir, A.

1987-01-01

The ASSERT subchannel code has been developed specifically to model flow and phase distributions within CANDU fuel bundles. ASSERT uses a drift-flux model that permits the phases to have unequal velocities, and can thus model phase separation tendencies that may occur in horizontal flow. The basic principles of ASSERT are outlined, and computed results are compared against data from various experiments for validation purposes. The paper concludes with an example of the use of the code to predict critical heat flux in CANDU geometries

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

Assessment of boiling transition analysis code against data from NUPEC BWR full-size fine-mesh bundle tests

International Nuclear Information System (INIS)

Utsuno, Hideaki; Ishida, Naoyuki; Masuhara, Yasuhiro; Kasahara, Fumio

2004-01-01

Transient BT analysis code TCAPE based on mechanistic methods coupled with subchannel analysis has been developed for the evaluation on fuel integrity under abnormal operations in BWR. TCAPE consisted mainly of the drift-flux model, the cross-flow model, the film model and the heat transfer model. Assessment of TCAPE has been performed against data from BWR full-size fine-mesh bundle tests (BFBT), which consisted of two major parts: the void distribution measurement and the critical power measurement. Code and data comparison was made for void distributions with varying number of unheated rods in simulated actual fuel assembly. Prediction of steady-state critical power was compared with the measurement on full-scale bundle under a range of BWR operational conditions. Although the cross-sectional averaged void fraction was underestimated when it became lower, the accuracy was obtained that the averaged ratio 0.910 and its standard deviation 0.076. The prediction of steady-state critical power agreed well with the data in the range of BWR operations, where the prediction accuracy was obtained that the averaged ratio 0.997 and its standard deviation 0.043. These results demonstrated that TCAPE is well capable to predict two-phase flow distribution and liquid film dryout phenomena occurring in BWR rod bundles. Part of NUPEC BFBT database will be made available for an international benchmark exercise. The code assessment shall be continued against the OECD/NRC benchmark based on BFBT database. (author)

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)

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Two-phase turbulent mixing and buoyancy drift in rod bundles

International Nuclear Information System (INIS)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

International Nuclear Information System (INIS)

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

1976-07-01

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

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Experimental studies of the effect of functional spacers to annular flow in subchannels of a BWR fuel element

Energy Technology Data Exchange (ETDEWEB)

Damsohn, M., E-mail: damsohn@lke.mavt.ethz.c [ETH Zurich, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zuerich (Switzerland); Prasser, H.-M. [ETH Zurich, Department of Mechanical and Process Engineering, Sonneggstrasse 3, 8092 Zuerich (Switzerland)

2010-10-15

For the prediction of dryout in fuel elements of boiling water reactors, the dynamic behavior of the water film covering the fuel rod has to be understood. This paper provides high resolved experimental data of the liquid film and gives insight into the dynamic film behavior. The experiments of this work were conducted in a vertical channel representing a pair of neighboring subchannels of a BWR fuel rod bundle. Air and water at ambient pressure and temperature are used as model fluids, creating an annular flow in the test section. The influence of different functional spacer shapes on the liquid film has been studied. The heart of the instrumentation is a liquid film sensor (LFS), which measures the film thickness distribution around a half cylinder with a matrix of 64 x 16 measuring points with a time resolution of 10,000 frames per second and a spatial resolution of 2 mm x 2 mm. The high resolution allows for a visualization of the dynamic liquid film as a movie animation. Principals of the dynamic behavior of the liquid film are observed. The time-averaged film thickness distributions show that the spacers structure the liquid film significantly. The gaseous phase is accelerated due to the cross-section blockage caused by the spacer. This leads to a local thinning of the liquid film downstream of the spacer. Two statistical evaluation methods are presented to determine different dynamic wave properties: The wave velocity as function of the wave height, the traveling path of the waves and the location of wave separation and merge events. The first evaluation method shows that big waves move generally faster than small waves. The analysis further shows wave acceleration in close proximity of the spacer with subsequent deceleration further downstream. Analyzing the wave as a two-dimensional entity it can be seen that the wave paths are clearly structured by the spacer and hence do not travel circumferentially around the fuel rod. Wave separation and merge has a

Experimental studies of the effect of functional spacers to annular flow in subchannels of a BWR fuel element

International Nuclear Information System (INIS)

Damsohn, M.; Prasser, H.-M.

2010-01-01

For the prediction of dryout in fuel elements of boiling water reactors, the dynamic behavior of the water film covering the fuel rod has to be understood. This paper provides high resolved experimental data of the liquid film and gives insight into the dynamic film behavior. The experiments of this work were conducted in a vertical channel representing a pair of neighboring subchannels of a BWR fuel rod bundle. Air and water at ambient pressure and temperature are used as model fluids, creating an annular flow in the test section. The influence of different functional spacer shapes on the liquid film has been studied. The heart of the instrumentation is a liquid film sensor (LFS), which measures the film thickness distribution around a half cylinder with a matrix of 64 x 16 measuring points with a time resolution of 10,000 frames per second and a spatial resolution of 2 mm x 2 mm. The high resolution allows for a visualization of the dynamic liquid film as a movie animation. Principals of the dynamic behavior of the liquid film are observed. The time-averaged film thickness distributions show that the spacers structure the liquid film significantly. The gaseous phase is accelerated due to the cross-section blockage caused by the spacer. This leads to a local thinning of the liquid film downstream of the spacer. Two statistical evaluation methods are presented to determine different dynamic wave properties: The wave velocity as function of the wave height, the traveling path of the waves and the location of wave separation and merge events. The first evaluation method shows that big waves move generally faster than small waves. The analysis further shows wave acceleration in close proximity of the spacer with subsequent deceleration further downstream. Analyzing the wave as a two-dimensional entity it can be seen that the wave paths are clearly structured by the spacer and hence do not travel circumferentially around the fuel rod. Wave separation and merge has a

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

International Nuclear Information System (INIS)

Shepherd, I.M.

1982-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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.)

Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

International Nuclear Information System (INIS)

Cheng, Z.; Rao, Y.F.; Waddington, G.M.

2014-01-01

Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles

Multi-dimensional boron transport modeling in subchannel approach: Part I. Model selection, implementation and verification of COBRA-TF boron tracking model

Energy Technology Data Exchange (ETDEWEB)

Ozdemir, Ozkan Emre, E-mail: ozdemir@psu.edu [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Avramova, Maria N., E-mail: mna109@psu.edu [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802 (United States); Sato, Kenya, E-mail: kenya_sato@mhi.co.jp [Mitsubishi Heavy Industries (MHI), Kobe (Japan)

2014-10-15

Highlights: ► Implementation of multidimensional boron transport model in a subchannel approach. ► Studies on cross flow mechanism, heat transfer and lateral pressure drop effects. ► Verification of the implemented model via code-to-code comparison with CFD code. - Abstract: The risk of reflux condensation especially during a Small Break Loss Of Coolant Accident (SB-LOCA) and the complications of tracking the boron concentration experimentally inside the primary coolant system have stimulated and subsequently have been a focus of many computational studies on boron tracking simulations in nuclear reactors. This paper presents the development and implementation of a multidimensional boron transport model with Modified Godunov Scheme within a thermal-hydraulic code based on a subchannel approach. The cross flow mechanism in multiple-subchannel rod bundle geometry as well as the heat transfer and lateral pressure drop effects are considered in the performed studies on simulations of deboration and boration cases. The Pennsylvania State University (PSU) version of the COBRA-TF (CTF) code was chosen for the implementation of three different boron tracking models: First Order Accurate Upwind Difference Scheme, Second Order Accurate Godunov Scheme, and Modified Godunov Scheme. Based on the performed nodalization sensitivity studies, the Modified Godunov Scheme approach with a physical diffusion term was determined to provide the best solution in terms of precision and accuracy. As a part of the verification and validation activities, a code-to-code comparison was carried out with the STAR-CD computational fluid dynamics (CFD) code and presented here. The objective of this study was two-fold: (1) to verify the accuracy of the newly developed CTF boron tracking model against CFD calculations; and (2) to investigate its numerical advantages as compared to other thermal-hydraulics codes.

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

International Nuclear Information System (INIS)

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

1985-01-01

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

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

Science.gov (United States)

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-04-01

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

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

International Nuclear Information System (INIS)

Lee, M.

2000-01-01

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

Pressure drop variation as a function of axial and radial power distribution in CANDU fuel channel with standard and CANFLEX 43 bundles

International Nuclear Information System (INIS)

Catana, Alexandru; Department of Energy Danila, Nicolae; Prisecaru, Ilie; Dupleac, Daniel

2007-01-01

CANDU 600 nuclear reactors are usually fuelled with STANDARD (STD), 37 rods fuel bundles. Natural uranium (NU) dioxide (UO 2 ), is used as fuel composition. A new fuel bundle geometry called CANFLEX (CFX) with 43 rods is proposed and some new fuel composition are considered. Flexibility is the key word for the attempt to use some different fuel geometries and compositions for CANDU 600 nuclear reactors as well as for innovative ACR-700/1000 nuclear reactors. The fuel bundle considered in this paper is CFX-RU-0.90 that encodes the CANFLEX geometry, recycled dioxide uranium (RU) with 0.90% enrichment. The goal of this proposal is ambitious: a higher average discharge burn-up up to 14000 MWd/tU and, for the same amount of generated electric power, reduction in nuclear fuel fabrication, reduction of spent nuclear fuel radioactive waste and reduction of refueling operational work by using fewer bundles. An improved sub-channel approach for thermal-hydraulic analysis is used in this paper to compute some flow parameters, mainly the pressure drop along the CANDU 600 fuel channel when STD or CFX-RU-0.90 fuel bundles. Also an intermediate CFX-NU fuel bundle are used, for gradual comparison. For CFX-RU- 0.90 four fuel bundle shift refueling scheme is used instead of eight, that will determine different axial power distributions. At the same time radial power distribution is affected by the geometry and by the fuel composition of fuel bundle type used. Some other thermal-hydraulic flow parameters will be influenced, too. One of the most important parameter is pressure drop (PD) along the fuel channel because of its importance in drag force evaluation. We start with an axial power distribution, which is characteristic for a refueling scheme of eight or four fuel bundles on a shift. Comparative results are presented between STD37, CFX-NU CFX-RU-0.90 fuel bundles in a CANDU nuclear reactor operating conditions. Neutron flux distribution analysis shows that four bundle shift

Description and validation of ANTEO, an optimised PC code the thermalhydraulic analysis of fuel bundles

International Nuclear Information System (INIS)

Cevolani, S.

1995-01-01

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 such a code was made possible by two facts: firstly, the increase, in the computing power of the desk machines; secondly, the fact that 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. (author)

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

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

Measurement of blockage in deformed LWR multi-rod arrays

International Nuclear Information System (INIS)

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

1983-01-01

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

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

International Nuclear Information System (INIS)

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

1976-01-01

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

Heat removal in gas-cooled fuel rod clusters

International Nuclear Information System (INIS)

Rehme, K.

1975-01-01

For a thermo- and fluid-dynamic analysis of fuel rod cluster subchannels for gas-cooled breeder reactors, the following values must be verified: a) friction coefficient as flow parameter; b) Stanton number as heat transfer parameter; c) influence of spacers on friction coefficient and Stanton number; d) heat and mass exchange between subchannels with different temperatures. These parameters are established by combining results of single experiments and of integral experiments. Mention is made of further studies to be performed in order to determine the heat removal from gas-cooled fast breeder fuel elements. (HR) [de

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-15

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

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

International Nuclear Information System (INIS)

Araya, Fumimasa; Nakatsuka, Toru; Yoritsune, Tsutomu

2002-10-01

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

Development and application of a novel technique for the measurement of mixing between subchannels

International Nuclear Information System (INIS)

Silin, Nicolas

2004-01-01

In this thesis we present the development of an experimental method for the measurement of mixing between coupled subchannels through the use of thermal traces.As this method can be applied to compact heterogeneous subchannels with high water flows and with presence of inserts and appendages, it is specially suited for the development of nuclear fuel elements, while showing advantages over other mixing measurement methods.The development of the method included the conceptual analysis of feasibility and application frame.Then the components necessary for the application of the technique to an experimental rig were developed and constructed, the most relevant being the high heat flux superficial heaters and a robust, low intrusivity, and a temperature measurement system with a precision better than 3mK.Preliminary tests were carried out to verify the technique, these included sensibility studies to flow rate and input power changes, settling time measurements, long term stability measurements and so forth. Also different error sources and their relative importance were analyzed.First, the method was applied to a channel of annular flow and then to a channel with three parallel rods generating four subchannels.Latter, measurements of inter subchannel mixing and mixing promoter performance assessment were carried out.The method developed allowed the proper measurement of the main parameters related to mixing, showing great potential as a design tool for nuclear fuel elements

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

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

Directory of Open Access Journals (Sweden)

A. A. Dunaitsev

2017-01-01

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

CANDU fuel bundle deformation modelling with COMSOL multiphysics

International Nuclear Information System (INIS)

Bell, J.S.; Lewis, B.J.

2012-01-01

Highlights: ► The deformation behaviour of a CANDU fuel bundle was modelled. ► The model has been developed on a commercial finite-element platform. ► Pellet/sheath interaction and end-plate restraint effects were considered. ► The model was benchmarked against the BOW code and a variable-load experiment. - Abstract: A model to describe deformation behaviour of a CANDU 37-element bundle has been developed under the COMSOL Multiphysics finite-element platform. Beam elements were applied to the fuel elements (composed of fuel sheaths and pellets) and endplates in order to calculate the bowing behaviour of the fuel elements. This model is important to help assess bundle-deformation phenomena, which may lead to more restrictive coolant flow through the sub-channels of the horizontally oriented bundle. The bundle model was compared to the BOW code for the occurrence of a dry-out patch, and benchmarked against an out-reactor experiment with a variable load on an outer fuel element.

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

International Nuclear Information System (INIS)

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

1992-10-01

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

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

Science.gov (United States)

Najeeb, Umair

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

Computer simulation of thermal-hydraulic transient events in multi-circuits with multipumps; Simulacao computacional de eventos termo-hidraulicos transitorios em multicircuitos com multibombas

Energy Technology Data Exchange (ETDEWEB)

Veloso, Marcelo Antonio

2003-07-01

PANTERA-2 (from Programa para Analise Termo-hidraulica de Reatores a Agua - Program for Thermal-hydraulic Analysis of Water Reactors, Version 2), whose fundamentals are described in this work, is intended to carry out rod bundle subchannel analysis in conjunction with multiloop simulation. It solves simultaneously the conservation equations of mass, axial and lateral momentum, and energy for subchannel geometry coupled with the balance equations that describe the fluid flows in any number of coolant loops connected to a pressure vessel containing the rod bundle. As far as subchannel analysis is concerned, the basic computational strategy of PANTERA-2 comes from COBRA codes, but an alternative implicit solution method oriented to the pressure field has been used to solve the finite difference approximations for the balance laws. The results provided by the subchannel model comprise the fluid density, enthalpy, flow rate, and pressure fields in the subchannels. The loop model predicts the individual loop flows, total flow through the pressure vessel, and pump rotational speeds as a function of time subsequent to the failure of any number of the coolant pumps. The flow transients in the loops may initiated by partial, total or sequential loss of electric power to the operating pumps. Transient events caused by either shaft break or rotor locking may also be simulated. The changes in rotational speed of the pumps as a function of rime are determined from a torque balance. Pump dynamic head and hydraulic torque are calculated as a function of rotational speed and volumetric flow from two polar homologous curves supplied to the code in the tabular form. In order to illustrate the analytical capability of PANTERA-2, three sample problems are presented and discussed. Comparisons between calculated and measured results indicate that the program reproduces with a good accuracy experimental data for subchannel exit temperatures and critical heat fluxes in 5x5 rod bundles. It

Computer simulation of thermal-hydraulic transient events in multi-circuits with multipumps

International Nuclear Information System (INIS)

Veloso, Marcelo Antonio

2003-01-01

PANTERA-2 (from Programa para Analise Termo-hidraulica de Reatores a Agua - Program for Thermal-hydraulic Analysis of Water Reactors, Version 2), whose fundamentals are described in this work, is intended to carry out rod bundle subchannel analysis in conjunction with multiloop simulation. It solves simultaneously the conservation equations of mass, axial and lateral momentum, and energy for subchannel geometry coupled with the balance equations that describe the fluid flows in any number of coolant loops connected to a pressure vessel containing the rod bundle. As far as subchannel analysis is concerned, the basic computational strategy of PANTERA-2 comes from COBRA codes, but an alternative implicit solution method oriented to the pressure field has been used to solve the finite difference approximations for the balance laws. The results provided by the subchannel model comprise the fluid density, enthalpy, flow rate, and pressure fields in the subchannels. The loop model predicts the individual loop flows, total flow through the pressure vessel, and pump rotational speeds as a function of time subsequent to the failure of any number of the coolant pumps. The flow transients in the loops may initiated by partial, total or sequential loss of electric power to the operating pumps. Transient events caused by either shaft break or rotor locking may also be simulated. The changes in rotational speed of the pumps as a function of rime are determined from a torque balance. Pump dynamic head and hydraulic torque are calculated as a function of rotational speed and volumetric flow from two polar homologous curves supplied to the code in the tabular form. In order to illustrate the analytical capability of PANTERA-2, three sample problems are presented and discussed. Comparisons between calculated and measured results indicate that the program reproduces with a good accuracy experimental data for subchannel exit temperatures and critical heat fluxes in 5x5 rod bundles. It

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

Braz Filho, F.A.

1987-01-01

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

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Analysis of CO2 draining in the supercritical subchannel of an MMR reactor using CFD techniques

International Nuclear Information System (INIS)

Dutra, Carolina S.B.; Ribeiro, Felipe P.; Su, Jian

2017-01-01

This work aims to analyze the steady state thermal hydraulic behavior of the supercritical CO 2 by means of the computational modeling of the triangular arrangement subchannel of the KAIST MMR (Korea Advanced Institute of Science and Technology). The mathematical model is composed of Navier-Stokes equations of Reynolds average (RANS), with the model of turbulence k-omega SST (Shear Stress Transport) for the fluid refrigerant in the subchannel and the heat conduction equation for the fuel and coating. The thermodynamic properties of CO 2 are implemented from National Institute of Standards and Technology (NIST) data. The transport equations were solved using a commercial CFD (Fluid Dynamics) tool, ANSYS FLUENT. Initially, a geometric model was constructed, using the ICEM software, composed of fuel, interstice, coating, refrigerant and two adiabatic regions of the fuel rod. The numerical solution used was validated by simulation of a simpler vertical circular tube and by comparisons with numerical and experimental results available in the literature. Using these results, we performed the mesh convergence study and initialized the simulations in the subchannel

The Analysis for the Effect of Mixing Vane Shape on TDC

International Nuclear Information System (INIS)

Moon, Kang Hoon; Park, Ho Young; Kim, Kang Hoon; Park, Eung Jun; Suh, Jung Min

2011-01-01

The Thermal Diffusion Coefficient (TDC) is an input parameter to subchannel code, and it is required to predict local flow conditions in a PWR fuel bundle. TDC influences on the prediction of thermal interchange or mixing of thermal energy between the hot subchannel and interconnected adjacent subchannels. The thermal mixing term in the energy equation is generally represented in terms of a non-dimensional inverse Peclet number or TDC. The parameters associated with thermal mixing can be defined as Eq.(1): TDC =Pe x De/a (1) where: Pe : Inverse Peclet Number (dimensionless) = ε/ Va De : Equivalent hydraulic diameter, in. a : Lateral flow area between channels per unit length, in 2 /in ε : Mixing coefficient, in 2 /sec V : Velocity, in/sec TDC is an important factor to evaluate thermal performance. So, flow temperature maps were obtained from the 5x5 rod bundle test section to assess the thermal performance of corresponding fuels. The flow temperatures were measured by thermocouple at the end of heated length and the centroid of subchannel. There are two typical methods to arrange the hot and cold fuel rods as shown in Fig. 1. Configuration Fig. 1(b) is adopted in this work. This paper presents how to determine the TDC and verifies whether all TDC with the effect of mixing vane shape is valid with respect to current design value

REBEKA bundle experiments

International Nuclear Information System (INIS)

Wiehr, K.

1988-05-01

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

Development of generalized boiling transition model applicable for wide variety of fuel bundle geometries. Basic strategy and numerical approaches

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)

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

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

A survey of blockage measurement methods used in PWR multi-rod experiments

Energy Technology Data Exchange (ETDEWEB)

Hindle, E.D.; Jones, C.; Whitty, S. (AEA Reactor Services, Springfield (UK))

1986-05-01

The deformation characteristics of Zircaloy multi-rod arrays are being investigated in laboratory and in-reactor tests, and heat transfer experiments are being carried out on pre-deformed arrays. The primary objective is to demonstrate that cladding distension occurring under hypothetical loss-of-coolant accident (LOCA) conditions will not impede the PWR emergency coolant flow during the reflood stage to the extent that unacceptably high cladding temperatures are reached, i.e. that a coolable geometry is maintained. This Report critically reviews the current methods for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed even though it still has drawbacks in the case of overall blockage measurement. A method for automatically measuring the individual rod strain, general cluster blockage sub-channel blockage and sub-channel perimeter changes is described and the results from a deformed array presented. (author).

A survey of blockage measurement methods used in PWR multi-rod experiments

International Nuclear Information System (INIS)

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

1986-05-01

The deformation characteristics of Zircaloy multi-rod arrays are being investigated in laboratory and in-reactor tests, and heat transfer experiments are being carried out on pre-deformed arrays. The primary objective is to demonstrate that cladding distension occurring under hypothetical loss-of-coolant accident (LOCA) conditions will not impede the PWR emergency coolant flow during the reflood stage to the extent that unacceptably high cladding temperatures are reached, i.e. that a coolable geometry is maintained. This Report critically reviews the current methods for measuring blockage in multi-rod arrays and discusses their application. A new definition which overcomes the deficiencies of the previous methods is proposed even though it still has drawbacks in the case of overall blockage measurement. A method for automatically measuring the individual rod strain, general cluster blockage sub-channel blockage and sub-channel perimeter changes is described and the results from a deformed array presented. (author)

CANAL user's manual

International Nuclear Information System (INIS)

Faya, A.; Wolf, L.; Todreas, N.

1979-11-01

CANAL is a subchannel computer program for the steady-state and transient thermal hydraulic analysis of BWR fuel rod bundles. The purpose of this manual is to introduce the user into the mechanism of running the code by providing information about the input data and options

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

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

Science.gov (United States)

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

2012-11-01

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

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

International Nuclear Information System (INIS)

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

1981-07-01

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

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)

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

International Nuclear Information System (INIS)

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

1976-01-01

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

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

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

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

International Nuclear Information System (INIS)

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

2002-04-01

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

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.

COMETHE III-M for transient fuel rod behaviour prediction

International Nuclear Information System (INIS)

Billaux, M.; Vliet, J. van

1983-01-01

The COMETHE III-M version is being developed in order to provide fuel rod behaviour prediction capability both in steady-state and in transient situations. It also allows to estimate the fuel rod enthalpy evolution versus time or burnup which may be important in core-related safety studies. This paper describes the transient heat transfer models, including transient heat conduction inside the fuel rod, and a subchannel model providing transient flow as well as enthalpy calculation capability. Transient fission gas release is also modelled on basis of the change rate of oxide temperature. The models are illustrated by a few calculation examples. (author)

Flow distribution in adjacent subchannels of unequal size

International Nuclear Information System (INIS)

Bugg, J.D.

1985-11-01

This report describes an experimental and analytic investigation of the single phase flow distribution in subchannel geometries. It was intended as an investigation of fundamental transport mechanisms and therefore concentrated on simple geometries with two interconnected subchannels. The experimental phase consisted of detailed measuremnts of the fluid velocity in a geometry representing two communicating subchannels of different sizes. These measurements were made at three axial locations along the test section. The size of one of the subchannels was varied to give subchannel area ratios of 1.0, 0.68, 0.50 and 0.32. Two Reynolds numbers (108000 and 180000) were investigated. Axial pressure gradient data for all of these cases was also taken. The analytic phase concentrated on applying a three dimensional finite difference fluid flow code to subchannel geomtries. The code was applied to the cases studied in the experiment as well as other investigator's results. It used the two equation K-ε turbulence model. The performance of this model was assessed. Unique features of the subchannel flows were identified and discussed. Conclusions regarding the transport mechanisms involved and the ability of a multidimensional code to predict the flow fields reliably were presented

Measurement of droplet dynamics across grid spacer in mist cooling of subchannel of PWR

International Nuclear Information System (INIS)

Lee, S.L.; Sheen, H.J.; Cho, S.K.; Issapour, I.

1984-01-01

An experiment was conducted of the dynamics and heat transfer of a droplet-vapor mist flow across a test grid spacer in a flow channel of 2 x 2 electrically heated simulation fuel rods. Embedded thermocouples were used to measure the rod cladding temperature and an unshielded Chromel-Alumel thermocouple was transversed in the center of the subchannel to measure the temperature of the water and steam coolant phases at various axial locations. Thermocouples were also embedded in the test grid spacer. Optical measurements of the size and velocity distributions of droplets and the velocity distribution of the superheated steam were made by special laser-Doppler anemometry techniques through quartz glass windows immediately upstream and downstream of the test grid spacer. Experiments over a range of steam and injected water flow rates and rod heat flux have been performed and some representative results and discussions are presented

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-05-15

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

COBRA-3M: a digital computer code for analyzing thermal-hydraulic behavior in pin bundles

International Nuclear Information System (INIS)

Marr, W.W.

1975-03-01

The COBRA-3M computer program is a modification of the thermal-hydraulic subchannel-analysis program COBRA-III. It includes detailed thermal models of fuel pin and duct wall. It is especially suitable for analyzing small pin bundles used in in-reactor or out-of-reactor experiments. (U.S.)

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Detection of flow mixing processes using transmission methods in high-duty heat exchanging apparatus

International Nuclear Information System (INIS)

Seiffert, V.

1981-01-01

The COBRA-IIIC program modified by MIT has been further improved for verifying the experimental studies described in the thesis. This work has been accompanied by a review and modification of the relevant analytical equations. A mathematical relationship has been set up for the cross-mixing phenomenon of shearing flow in narrowest cross-section between two heating rods, the relationship being taken into account in the sub-channel analysis. Despite the very complex and superposing processes of the problem studied, the results obtained by the improved sub-channel analysis program using the nearly derived cross-mixing approach are quantitatively well confirmed by comparison with experimental data. Applying the improved sub-channel analysis program to describing the author's two-phase flow experiments (air-water and water-steam) with rod bundle geometries to be found in the literature, the cross-mixing approach presented in the thesis is shown to be reliable (orig./GL) [de

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)